Advanced Tokamak Research Articles

Comparison of Options for a Pilot Plant Fusion Nuclear Mission

A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS).With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.

Mission and Readiness Assessment for Fusion Nuclear Facilities

Magnetic fusion development toward DEMO will most likely require a number of fusion nuclear facilities (FNF), intermediate between ITER and DEMO, to test and validate plasma and nuclear technologies and to advance the level of system integration. The FNF mission space is wide, ranging from basic materials research to net electricity demonstration, so there is correspondingly a choice among machine options, scope, and risk in planning such a step. Readiness requirements to proceed with a DEMO are examined, and two FNF options are assessed in terms of the contributions they would make to closing DEMO readiness gaps, and their readiness to themselves proceed with engineering design about ten years from now. An advanced tokamak (AT) pilot plant with superconducting coils and a mission to demonstrate net electricity generation would go a long way toward DEMO. As a next step, however, a pilot plant would entail greater risk than a copper-coil FNSF-AT with its more focussed mission and technology requirements. The stellarator path to DEMO is briefly discussed. Regardless of the choice of FNF option, an accompanying science and technology development program, also aimed at DEMO readiness, is absolutely essential.

ARIES-ACT1 Power Core Engineering

ARIES-ACT1 is an advanced tokamak power plant conceptual design that utilizes SiC composite structural material in the blanket and PbLi as the tritium breeder and coolant. This design concept represents an evolutionary step from ARIES-AT, which has guided tokamak research programs for the past decade. In conjunction with a helium Brayton power cycle, the high primary coolant outlet temperature allows thermal conversion efficiency of 58%. The self-cooled blanket and He-cooled W-alloy divertor provide the ability to survive relatively high power density with acceptable projected lifetime. In ARIES-ACT1, we attempted to add “robustness” to the design point without major sacrifices in performance. In this paper, we will discuss the main features of the power core and selected details in the design and analysis.

Modelling of hybrid scenario: from present-day experiments towards ITER

The ‘hybrid’ scenario is an attractive operating scenario for ITER since it combines long plasma duration with the reliability of the reference H-mode regime. We review the recent European modelling effort carried out within the Integrated Scenario Modelling group which aims at (i) understanding the underlying physics of the hybrid regime in ASDEX-Upgrade and JET and (ii) extrapolating them towards ITER. JET and ASDEX-Upgrade hybrid scenarios performed under different experimental conditions have been simulated in an interpretative and predictive way in order to address the current profile dynamics and its link with core confinement, the relative importance of magnetic shear, s, and E × B flow shear on the core turbulence, pedestal stability and H–L transition. The correlation of the improved confinement with an increased s/q at outer radii observed in JET and ASDEX-Upgrade discharges is consistent with the predictions based on the GLF23 model applied in the simulations of the ion and electron kinetic profiles. Projections to ITER hybrid scenarios have been carried out focusing on optimization of the heating/current drive schemes to reach and ultimately control the desired plasma equilibrium using ITER actuators. Firstly, access condition to the hybrid-like q-profiles during the current ramp–up phase has been investigated. Secondly, from the interpreted role of the s/q ratio, ITER hybrid scenario flat-top performance has been optimized through tailoring the q-profile shape and pedestal conditions. EPED predictions of pedestal pressure and width have been used as constraints in the interpretative modelling while the core heat transport is predicted by GLF23. Finally, model-based approach for real-time control of advanced tokamak scenarios has been applied to ITER hybrid regime for simultaneous magnetic and kinetic profile control.

Integrated magnetic and kinetic control of advanced tokamak plasmas on DIII-D based on data-driven models

The first real-time profile control experiments integrating magnetic and kinetic variables were performed on DIII-D in view of regulating and extrapolating advanced tokamak scenarios to steady-state devices and burning plasma experiments. Device-specific, control-oriented models were obtained from experimental data using a generic two-time-scale method that was validated on JET, JT-60U and DIII-D under the framework of the International Tokamak Physics Activity for Integrated Operation Scenarios (Moreau et al 2011 Nucl. Fusion 51 063009). On DIII-D, these data-driven models were used to synthesize integrated magnetic and kinetic profile controllers. The neutral beam injection (NBI), electron cyclotron current drive (ECCD) systems and ohmic coil provided the heating and current drive (H&CD) sources. The first control actuator was the plasma surface loop voltage (i.e. the ohmic coil), and the available beamlines and gyrotrons were grouped to form five additional H&CD actuators: co-current on-axis NBI, co-current off-axis NBI, counter-current NBI, balanced NBI and total ECCD power from all gyrotrons (with off-axis current deposition). Successful closed-loop experiments showing the control of (a) the poloidal flux profile, Ψ(x), (b) the poloidal flux profile together with the normalized pressure parameter, βN, and (c) the inverse of the safety factor profile, , are described.

A fast, magnetics-free flux surface estimation and q-profile reconstruction algorithm for feedback control of plasma profiles

The flux surfaces' layout and the magnetic winding number q are important quantities for the performance and stability of tokamak plasmas. Normally, these quantities are iteratively derived by solving the plasma equilibrium for the poloidal and toroidal flux.In this work, a fast, non-iterative and magnetics-free numerical method is proposed to estimate the shape of the flux surfaces by an inward propagation of the plasma boundary shape, as can be determined for example by optical boundary reconstruction described in Hommen (2010 Rev. Sci. Instrum. 81 113504), toward the magnetic axis, as can be determined independently with the motional Stark effect (MSE) diagnostic. Flux surfaces are estimated for various plasma regimes in the ITER, JET and MAST tokamaks and are compared with results of CRONOS reconstructions and simulations, showing agreement to within 1% of the minor radius for almost all treated plasmas.The availability of the flux surface shapes combined with the pitch angles measured using MSE allow the reconstruction of the plasma q-profile, by evaluating the contour-integral over the flux surfaces of the magnetic field pitch angle. This method provides a direct and exact measure of the q-profile for arbitrary flux surface shapes, which does not rely on magnetic measurements. Results based on estimated flux surface shapes show agreement with CRONOS q-profiles of better than 10%. The impact of the shape of the flux surfaces on the q-profile, particularly the profiles of elongation and Shafranov shift, and offsets in plasma boundary and the magnetic axis are assessed.OFIT+ was conceived for real-time plasma profile control experiments and advanced tokamak operation, and provides quickly and reliably the mapping of actuators and sensors to the minor radius as well as the plasma q-profile, independent of magnetic measurements.

Progress of KSTAR 5-GHz Lower Hybrid Current Drive System

A 5-GHz steady-state lower hybrid (LH) current drive (LHCD) system is planned to support steady-state and advanced tokamak operation on the Korea Superconducting Tokamak Advanced Research (KSTAR) experiment. As an initial phase, a pulsed 5-GHz, 500-kW LHCD system has been installed in KSTAR for basic experimental studies of the LH coupling and flux saving in the plasma current ramp-up, prior to long-pulse noninductive operation in KSTAR. A Toshiba-made klystron developed in collaboration with Pohang University of Science and Technology in 2006 is utilized for the initial KSTAR LHCD system. The LH launcher is designed as a fully active waveguide grill type with a parallel refractive index n[parallel] value ranging from 1.8 to 4.3 and with high directivity. In the initial stage, the LH launcher consists of eight columns of four-way power splitters and two columns of dummy waveguides, one on each side. The operational n[parallel] value is fixed at 2.1 but can be adjusted by replacing waveguide components external to the vacuum vessel. Since the target operation pulse duration of the initial LHCD system is 2 s with an output power of 500 kW at the klystron window, the prototype klystron was recently successfully conditioned to a radio frequency power of 514 kW for a maximum pulse duration of 3 s using a matched dummy load (voltage standing wave ratio of 1.16:1). This paper presents the progress of the initial KSTAR LHCD system and the performance test results of the prototype klystron. The research plan aiming at steady-state LHCD operation in KSTAR is also described in this paper.

Toroidal current profile control during low confinement mode plasma discharges in DIII-D via first-principles-driven model-based robust control synthesis

In order for ITER to be capable of operating in advanced tokamak operating regimes, characterized by a high fusion gain, good plasma confinement, magnetohydrodynamic stability and a non-inductively driven plasma current, for extended periods of time, several challenging plasma control problems still need to be solved. Setting up a suitable toroidal current density profile in the tokamak is key for one possible advanced operating scenario characterized by non-inductive sustainment of the plasma current. At the DIII-D tokamak, the goal is to create the desired current profile during the ramp-up and early flat-top phases of the plasma discharge and then actively maintain this target profile for the remainder of the discharge. The evolution in time of the toroidal current profile in tokamaks is related to the evolution of the poloidal magnetic flux profile, which is modelled in normalized cylindrical coordinates using a first-principles, nonlinear, dynamic partial differential equation (PDE) referred to as the magnetic diffusion equation. The magnetic diffusion equation is combined with empirical correlations developed from physical observations and experimental data from DIII-D for the electron temperature, the plasma resistivity and the non-inductive current drive to develop a simplified, control-oriented, nonlinear, dynamic PDE model of the poloidal flux profile evolution valid for low confinement mode discharges. In this work, we synthesize a robust feedback controller to reject disturbances and track a desired reference trajectory of the poloidal magnetic flux gradient profile by employing the control-oriented model of the system. A singular value decomposition of the static gain matrix of the plant model is utilized to identify the most relevant control channels and is combined with the dynamic response of system around a given operating trajectory to design the feedback controller. A general framework for real-time feedforward + feedback control of magnetic and kinetic plasma profiles was implemented in the DIII-D Plasma Control System and was used to demonstrate the ability of the feedback controller to control the toroidal current profile evolution in the DIII-D tokamak. These experiments constitute the first time ever a first-principles-driven, model-based, closed-loop magnetic profile controller was successfully implemented and tested in a tokamak device.

Ideal MHD Stability Prediction and Required Power for EAST Advanced Scenario

The Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. The ideal magnetohydrodynamic (MHD) stability and required power for the EAST advanced tokamak (AT) scenario with negative central shear and double transport barrier (DTB) are investigated. With the equilibrium code TOQ and stability code GATO, the ideal MHD stability is analyzed. It is shown that a moderate ratio of edge transport barriers' (ETB) height to internal transport barriers' (ITBs) height is beneficial to ideal MHD stability. The normalized beta βN limit is about 2.20 (without wall) and 3.70 (with ideal wall). With the scaling law of energy confinement time, the required heating power for EAST AT scenario is calculated. The total heating power Pt increases as the toroidal magnetic field BT or the normalized beta βN is increased.

A new approach to plasma profile control in ITER

Advanced tokamak operation in ITER, such as the steady-state and hybrid modes, requires an active real-time feedback control of plasma profiles to achieve the advanced regimes for sustained operation. In this work, we have explored a potentially robust control technique that simplifies the active real-time control of electron temperature and safety factor profiles in ITER. As a new and simple approach, static responses of the plasma profiles to power changes of auxiliary heating and current drive are modelled and updated in real-time, differing from the techniques which use a dynamic model deduced from identification experiments, or even a simplified explicit model. To allow real-time update of the plasma profile response model, the underlying physics is simplified with several assumptions. The electron temperature profile response is modelled by simplifying the electron heat transport equation. The safety factor profile response is modelled by directly relating it to the changes of source current density profiles. The required actuator power changes are calculated using the singular value decomposition technique, taking the saturation of the actuator powers into account. The potential of this control technique has been tested by applying it to simulations of the ITER hybrid mode operation using CRONOS. In these simulations, the electron temperature and safety factor profiles were well controlled either independently or simultaneously.

Fusion Nuclear Science Facility-AT: A Material and Component Testing Device

A fusion nuclear science facility (FNSF) is a necessary complement to ITER, especially in the area of material and component testing needed for DEMO design development. FNSF-AT, which takes advantage of advanced tokamak (AT) physics, should have neutron wall loading of 1-2 MW/m2, continuous operation for periods of up to 2 weeks, a duty factor goal of 0.3 per year, and an accumulated fluence of 3-6 MW-yr/m2 (~30-60 dpa) in 10 years to enable the qualification of structural, blanket, and functional materials, components, and corresponding ancillary equipment necessary for the design and licensing of a DEMO. Base blankets with a ferritic steel structure and selected tritium blanket materials will be tested and used for the demonstration of tritium sufficiency. Additional test ports at the outboard midplane will be reserved for test blankets with advanced designs or exotic materials and electricity production for integrated high-fluence testing in a DT fusion spectrum. FNSF-AT will be designed using conservative implementations of all elements of AT physics to produce 150-300 MW of fusion power with modest energy gain (Q < 7) in a modest-sized normal conducting coil device. It will demonstrate and help to select the DEMO plasma-facing, structural, tritium-breeding, and functional materials and ancillary equipment including diagnostics. It will also demonstrate the necessary tritium fuel cycle, design and cooling of the first wall chamber, and divertor components. It will contribute to the knowledge on material qualification, licensing, operational safety, and remote maintenance necessary for DEMO design.

Determination of neutral beam energy fractions from collisional radiative measurements on DIII-D

Neutral beams based on positive ion source technology are a key component of contemporary fusion research. An accurate assessment of the injected beam species mix is important for determining the actual plasma heating and momentum input as well as proper interpretation of beam-based diagnostics. On DIII-D, the main ion charge-exchange spectroscopy system is used to extract well-resolved intensity ratios of the Doppler-shifted D(α) emission from the full, half, and third energy beam components for a variety of beam operational parameters. In conjunction with accurate collisional-radiative modeling, these measurements indicate the assumed species mix and power fractions can vary significantly and should be regularly monitored and updated for the most accurate interpretation of plasma performance. In addition, if stable active control of the power fractions can be achieved through appropriate source tuning, the resulting control over the deposition profile can serve as an additional experimental knob for advanced tokamak studies, e.g., varying the off axis beam current drive without altering the beam trajectory.

Physics of resistive wall modes

The advanced tokamak regime is a promising candidate for steady-state tokamak operation which is desirable for a fusion reactor. This regime is characterized by a high bootstrap current fraction and a flat or reversed safety factor profile, which leads to operation close to the pressure limit. At this limit, an external kink mode becomes unstable. This external kink is converted into the slowly growing resistive wall mode (RWM) by the presence of a conducting wall. Reduction of the growth rate allows one to act on the mode and to stabilize it. There are two main factors which determine the stability of the RWM. The first factor comes from external magnetic perturbations (error fields, resistive wall, feedback coils, etc). This part of RWM physics is the same for tokamaks and reversed field pinch configurations. The physics of this interaction is relatively well understood and based on classical electrodynamics. The second ingredient of RWM physics is the interaction of the mode with plasma flow and fast particles. These interactions are particularly important for tokamaks, which have higher plasma flow and stronger trapped particle effects. The influence of the fast particles will also be increasingly more important in ITER and DEMO which will have a large fraction of fusion born alpha particles. These interactions have kinetic origins which make the computations challenging since not only particles influence the mode, but also the mode acts on the particles. Correct prediction of the ‘plasma–RWM’ interaction is an important ingredient which has to be combined with external field's influence (resistive wall, error fields and feedback) to make reliable predictions for RWM behaviour in tokamaks. All these issues are reviewed in this paper.

Lower hybrid current drive at high density in the multi-pass regime

Assessing the performance of lower hybrid current drive (LHCD) at high density is critical for developing non-inductive current drive systems on future steady-state experiments. Excellent LHCD efficiency has been observed during fully non-inductive operation (η=2.0−2.5×1019 AW–1m–2 at n¯e=0.5×1020 m–3) on Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under conditions (ne, magnetic field and topology, and LHCD frequency) relevant to ITER [S. Shiraiwa et al., Nucl. Fusion 51, 103024 (2011)]. To extend these results to advanced tokamak regimes with higher bootstrap current fractions on C-Mod, it is necessary to increase n¯e to 1.0−1.5×1020 m−3. However, the number of current-carrying, non-thermal electrons generated by LHCD drops sharply in diverted configurations at densities that are well below the density limit previously observed on limited tokamaks. In these cases, changes in scrape off layer (SOL) ionization and density profiles are observed during LHCD, indicating that significant power is transferred from the LH waves to the SOL. Fokker-Planck simulations of these discharges utilizing ray tracing and full wave propagation codes indicate that LH waves in the high density, multi-pass absorption regime linger in the plasma edge, and SOL region, where absorption near or outside the LCFS results in the loss of current drive efficiency. Modeling predicts that non-thermal emission increases with stronger single-pass absorption. Experimental data show that increasing Te in high density LH discharges results in higher non-thermal electron emission, as predicted by the models.

Theoretical analysis of the ICRH antenna's impedance matching for ELMy plasmas on EAST

AbstractA well-optimized design of an ion cyclotron resonance heating (ICRH) antenna is very important for steady-state plasma heating with high radio frequency (RF) power of several tens of megawatts. However, a sharp decrease in the coupling RF power because of impedance mismatch of ICRH system is an issue that must be resolved for present-day fusion reactors and International Thermonuclear Experimental Reactor. This paper has theoretically analyzed the ICRH antenna's impedance matching for ELMy plasmas on experimental advanced superconducting tokamak (EAST) by the transmission line theory. The results indicate that judicious choice of the optimal feeder location is found useful for adjustable capacitors' tolerance to the variations of the antenna input impedance during edge-localized mode (ELM) discharge, which is expected to be good for the design of ICRH antenna system and for real-time feedback control during ELM discharge on EAST.

Key Features in the Operation of KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device is aimed at advanced tokamak (AT) research. Three years have passed since it achieved its first plasma in 2008. Because it is a superconducting machine and is working toward AT research, it has unique features in terms of the machine engineering and operation. The toroidal field (TF) magnet coils are made of Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn, which provide high TFs up to 3.5 T, and have been fully tested. The poloidal field (PF) magnet coils, consisting of both Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and NbTi, which have a maximum current of 25 kA in their design, were tested up to 15 kA. A thermal hydraulic analysis is being conducted for PF magnet coil operation. All plasma-facing components (PFCs) are equipped with water cooled graphite tiles and have the capability of being baked up to 350°C. A startup scenario, which considered both the effect of the ferromagnetic material in the cable in conduit conductor jacket of the magnet coils as well as a nonferromagnetic up-down asymmetry in the cryostat structure, was developed and demonstrated its effectiveness by the last two year's reliable operations. Passive stabilizers and in-vessel control coils (IVCCs) are key components to realize AT operation in KSTAR. The segmented IVCC coils were connected to form circular coils for internal vertical control in 2010, and diverted plasmas with high elongation (κ~1.8, δ>;0.6) were achieved. A neutral beam injection (NBI) system was developed aiming at 2 MW, 300 s per ion source which meets the long-pulse requirement of KSTAR. An NBI ion source with a power of 1.7 MW at 100 kV has been commissioned for 10 s. Finally, ELMy H-modes were successfully produced with 1.3-MW NBI power at a plasma current of 0.6 MA in the 2010 campaign. The first H-mode discharge (#4200) in KSTAR was achieved one year earlier than officially planned and was done at B <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">T</sub> = 2 T with I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">p</sub> = 0.6 MA in a well-balanced double null configuration after boronization on the PFC. Successful operations in the early days of KSTAR including H-mode experiments revealed the capability of advanced and steady-state operation which is essential for the international thermonuclear experimental reactor (ITER) and future fusion reactors.

Large ELM-like events triggered by core MHD in JET advanced tokamak plasmas: impact on plasmas profiles, plasma-facing components and heating systems

Large and infrequent collapse events have been observed in high βN advanced tokamak (AT) plasmas in JET. Although they have features similar to large ELMs, they were triggered by core MHD. They caused a considerable loss of the plasma thermal and fast particle energy (∼10% of the total stored energy), but the heat load in the divertor due to these collapse events was small as a fraction of the plasma energy loss compared with regular type-I ELMs. Instead, significant heating of the main chamber wall was observed. A large, toroidally asymmetric, increase in the neutral gas pressure outside the plasma was observed after such events, which caused arcs in the lower hybrid (LH) and ion cyclotron (IC) heating systems and increased reionization in the neutral beam (NB) injectors. The collapses resulted in a reduction in the electron and ion temperatures and toroidal rotation of the whole plasma, a rise in Zeff; and a sufficiently large increase in the peripheral electron density to completely black-out the ECE emission from the plasma core. These features have been modelled to gain an understanding of the plasma behaviour associated with these collapse events and the implication for the operation of AT plasma scenarios with high additional heating power will be discussed.

Robust Sliding Mode Control for Tokamaks

Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

The lower hybrid current drive system for steady-state operation of the Vulcan tokamak conceptual design

The steady-state current drive system for the Vulcan tokamak concept has been designed, taking into account requirements of high field, small size, and high operational wall temperature (B0=7T, R0=1.2m, Twall>800K). This lower hybrid current drive system allows steady-state operation by utilizing high field side launch, high RF source frequency (8GHz), and dedicated current drive ports. An iterative MHD and current drive solver is used to determine the ideal launching spectra and location to assure strong single pass absorption. It is found that with nominal Vulcan operational parameters (ne≈4×1020m−3, Te≈2.8keV, Ip=1.7MA, PLHCD=19.8MW) bootstrap currents of ∼70% and lower hybrid current drive efficiencies of 1.16×1019AWm−2 could be achieved. The optimized solution yielded advanced tokamak profiles with q values on-axis above 2. A conceptual design of the system is presented, which takes into account space, power, cooling, and launched spectrum requirements. The system is found to be compatible with the vacuum vessel design and requires cooling power of <1MW per waveguide bundle.

A Key to Improved Ion Core Confinement in the JET Tokamak: Ion Stiffness Mitigation due to Combined Plasma Rotation and Low Magnetic Shear

New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.