Advanced Tokamak Research Articles

A fusion development facility on the critical path to fusion energy

A fusion development facility (FDF) based on the tokamak approach with normal conducting magnetic field coils is presented. FDF is envisioned as a facility with the dual objective of carrying forward advanced tokamak (AT) physics and enabling the development of fusion energy applications. AT physics enables the design of a compact steady-state machine of moderate gain that can provide the neutron fluence required for FDF's nuclear science development objective. A compact device offers a uniquely viable path for research and development in closing the fusion fuel cycle because of the demand to consume only a moderate quantity of the limited supply of tritium fuel before the technology is in hand for breeding tritium.

An ITER-relevant passive active multijunction launcher for lower hybrid current drive in JET-grade plasmas

Lower hybrid current drive (LHCD) has been demonstrated to be an important actuator for controlling the current profiles of advanced tokamak plasmas, either in fully non-inductive or in hybrid regimes. To continue such scenario development in JET, and most notably towards extrapolations for ITER, an upgrade of its LH launcher is required to deliver more power (>5 MW) and to deal with a broad array of density situations at the plasma edge more consistently. These may range from steep gradients to ELM activity; extend from as low as the cut-off density nec to large and fast density variations and give rise to strong and rapid reflections of RF power. The passive active multijunction (PAM), advocated as the LH launcher concept for ITER, is claimed to be more resilient to such conditions, run with increased efficiency near nec and even with a small vacuum gap, being also specially appropriate for long pulse operation—all conditions hampering traditional designs. In addition, some of this leverage has been confirmed experimentally at the ITER-relevant FTU with the first PAM ever deployed, and more recently with Tore Supra's PAM. The JET version addressed in this study, with the latter focusing on the physics that drives and supports the adopted technical solutions, allows the injection of up to 5.2 MW at 3.7 GHz and N||peak = 1.9. With directivities better than 70%, reflection coefficients below 2% even for ∼nec and power densities comparable to those foreseen for ITER at 5 GHz, it would validate this concept in ITER-relevant scenarios.

Core transport properties in JT-60U and JET identity plasmas

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma shape, toroidal magnetic field ripple and dimensionless profiles as close as possible during the ITB triggering phase in terms of safety factor, normalized Larmor radius, normalized collision frequency, thermal beta, ratio of ion to electron temperatures. Similarities in the ITB triggering mechanisms and sustainment were observed when a good match was achieved of the most relevant normalized profiles except the toroidal Mach number. Similar thermal ion transport levels in the two devices have been measured in either monotonic or non-monotonic q-profiles. In contrast, differences between JET and JT-60U were observed on the electron thermal and particle confinement in reversed magnetic shear configurations. It was found that the larger shear reversal in the very centre (inside normalized radius of 0.2) of JT-60U plasmas allowed the sustainment of stronger electron density ITBs compared with JET. As a consequence of peaked density profile, the core bootstrap current density is more than five times higher in JT-60U compared with JET. Thanks to the bootstrap effect and the slightly broader neutral beam deposition, reversed magnetic shear configurations are self-sustained in JT-60U scenarios. Analyses of similarities and differences between the two devices address key questions on the validity of the usual assumptions made in ITER steady scenario modelling, e.g. a flat density profile in the core with thermal transport barrier? Such assumptions have consequences on the prediction of fusion performance, bootstrap current and on the sustainment of the scenario.

FAST: Feasibility Analysis for a Completely Superconducting Magnet System

FAST (Fusion Advanced Studies Torus), the Italian proposal for a European satellite facility to ITER, is a compact tokamak ( R <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">0</sub> = 1.82 m, a= 0.64 m, triangularity δ = 0.4) able to investigate non linear dynamics effects of alpha-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO (e.g.: full-tungsten wall and divertor and advanced liquid metal divertor). The machine is designed to operate with Deuterium plasmas in a dimensionless parameter range close to that of ITER and to access advanced tokamak regimes with long pulse duration with respect to the current diffusion time. It foresees a maximum magnetic field on plasma axis of 8.5 T and a maximum plasma current of 8 MA. In the present design phase, the feasibility of a superconducting solution for the magnet system is being investigated by ENEA. It consists of 18 Toroidal Field, 6 Poloidal Field and 6 Central Solenoid module coils, all of which wound by Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and NbTi Cable-In-Conduit Conductors. All the main aspects driving the magnets' design, from mechanical to neutronic and thermal analyses, are here presented and discussed.

Resistive wall mode stabilization by differential rotation in an analytic tokamak

It is well known that stabilization of the resistive wall mode (RWM) may allow fusion power to be significantly increased for a given magnetic field in advanced tokamak operation. The principle of stabilization of the RWM by rotation has been established both experimentally and theoretically. Recent experimental results have indicated stabilization of the RWM has been achieved with very small levels of rotation using balanced neutral beam injection. The framework of Connor et al (Connor et al 1988 Phys. Fluids 31 577) is used to develop two ideal plasma analytic toroidal models where stepped pressure profiles and careful ordering of terms are used to simplify the analysis. The first model has one resonant layer in the plasma and two resistive walls and the second has two resonant layers and one resistive wall. The RWM can be stabilized with slow rotation (∼0.5%ΩA, where ΩA is the Alfvénic rotation frequency) of a secondary resistive wall. A secordary rotating resistive wall can only stabilize the plasma if a perfectly conducting wall at that location would stabilize the plasma. Differential rotation in the plasma is investigated by rotating two resonant layers in the plasma at different rates. It is found in this model that differential rotation of the outer resonant surface can stabilize the RWM, with no rotation required at the inner surface. However, stabilization is not possible with the outer surface static and the inner surface rotating. Experimental rotation profiles for balanced neutral beam shots have higher levels of rotation near the plasma edge showing differential rotation between the outer surface and the wall may be part of the RWM stabilization mechanism.

Plasma models for real-time control of advanced tokamak scenarios

An integrated plasma profile control strategy, ARTAEMIS, is being developed for extrapolating present-day advanced tokamak (AT) scenarios to steady-state operation. The approach is based on semi-empirical modelling and was initially explored on JET (Moreau et al 2008 Nucl. Fusion 48 106001). This paper deals with the general applicability of this strategy for simultaneous magnetic and kinetic control on various tokamaks. The determination of the device-specific, control-oriented models that are needed to compute optimal controller matrices for a given operation scenario is discussed. The methodology is generic and can be applied to any device, with different sets of heating and current drive actuators, controlled variables and profiles. The system identification algorithms take advantage of the large ratio between the magnetic and thermal diffusion time scales and have been recently applied to both JT-60U and DIII-D data. On JT-60U, an existing series of high bootstrap current (∼70%), 0.9 MA non-inductive AT discharges was used. The actuators consisted of four groups of neutral beam injectors aimed at perpendicular injection (on-axis and off-axis), and co-current tangential injection (also on-axis and off-axis). On DIII-D, dedicated system identification experiments were carried out in the loop voltage (Vext) control mode (as opposed to current control) to avoid feedback in the response data from the primary circuit. The reference plasma state was that of a 0.9 MA AT scenario which had been optimized to combine non-inductive current fractions near unity with 3.5 < βN < 3.9, bootstrap current fractions larger than 65% and H98(y,2) = 1.5. Actuators other than Vext were co-current, counter-current and balanced neutral beam injection, and electron cyclotron current drive. Power and loop voltage modulations resulted in dynamic variations of the plasma current between 0.7 and 1.2 MA. It is concluded that the response of essential plasma parameter profiles to specific actuators of a given device can be satisfactorily identified from a small set of experiments. This provides, for control purposes, a readily available alternative to first-principles plasma modelling.

Self-suppression of double tearing modes via Alfvén resonance in rotating tokamak plasmas

Reversed magnetic shear configuration, a key method for improving plasma confinement in advanced tokamaks, is prone to exciting double tearing modes (DTMs) that can severely degrade the plasma confinement. In this letter, we reveal a new mechanism of suppressing the DTM instability due to the self-induced Alfvén resonance in rotating tokamak plasmas. The linear growth rate is reduced from ~SHp-1/3 of the fast DTM regime to ~SHp-3/5 of the slow single tearing mode regime, where SHp is magnetic Reynolds number. Instead of generating magnetic islands at the inner rational surface that can greatly enhance plasma transport in the core region, the formation of current sheets at resonance layers not only prevents the fast nonlinear DTM reconnection phase but also contributes to plasma heating.

The role of JET for the preparation of the ITER exploitation

The JET programme is devoted to the consolidation of ITER design choices and the qualification of ITER integrated regimes of operation. During the experimental campaigns carried out in 2008 and 2009 attention focussed on the test of the ITER-like ICRH antenna, the ITER scenario preparation, the verification of the adequacy of the ITER poloidal field coil design and the test of disruption mitigation methods such as massive gas injection. From 2011 the new ITER-like wall with all beryllium and tungsten plasma facing components, the neutral beam power upgrade and the enhanced control and diagnostic capability will allow key questions on plasma-wall interactions, fuel retention and plasma impurity control with the foreseen ITER wall materials to be addressed. Finally, feasibility studies have confirmed the option of installing an ITER-technology based 170 GHz/10 MW electron cyclotron resonance heating system for the control of MHD activity and the development of advanced tokamak scenarios, and 32 in-vessel coils for ELM control capable of producing magnetic perturbation spectra with a Chirikov parameter above unity for plasma currents up to 5 MA. During the ITER construction phase, JET will be the only device of its class in operation and will therefore play a key role in the preparation of ITER operations – saving time and reducing risk from the ITER programme.

Simserver simulation of a model-based current profile controller in the DIII-D Plasma Control System

Abstract Setting up a suitable current profile, characterized by a weakly reversed magnetic shear, has been demonstrated to be a key condition for one possible advanced tokamak operating scenario with improved confinement and possible steady-state operation. Experiments at DIII-D focus on creating the desired current profile during the plasma current ramp-up and early flat-top phases with the aim of maintaining this target profile throughout the subsequent phases of the discharge. The evolution in time of the current profile, or alternatively the safety factor q, is related to the evolution of the poloidal flux, which is modeled in normalized cylindrical coordinates using a partial differential equation referred to as the magnetic flux diffusion equation. A control-oriented model of the current profile evolution in DIII-D was recently developed for the plasma current ramp-up and early flat-top phases and used to synthesize both open-loop and closed-loop control schemes. In this work, we report on the implementation of an advanced model-based current profile controller in the DIII-D Plasma Control System (PCS) and on the assessment of this controller implementation in closed-loop Simserver (simulation server) simulations.

FAST: A European ITER satellite experiment in the view of DEMO

Fusion Advanced Studies Torus (FAST) aims to contribute to the exploitation of ITER and to explore innovative DEMO technology. FAST has been designed to study, in an integrated scenario: (a) relevant plasma-wall interaction problems, with a large power load ( P/ R ∼ 22 MW/m; P/ R 2 ∼ 12 MW/m 2) and with a full metallic wall; (b) to tackle operational problems in regimes with relevant fusion parameters; (c) to investigate the non-linear dynamics of fast particles (alpha like) in burning plasmas. FAST will operate on a wide parameters range, namely in high performance H-mode (BT ∼ 8.5 T; IP ∼ 8 MA) as well as in advanced Tokamak operation up to full non-inductive current scenario (IP ∼ 2 MA). The main heating is based on 30 MW ICRH, but the ports have been designed to allocate up to 20 MW of 1 MeV NNBI. Helium gas at 30 K is used for cooling of the full machine, a preliminary analysis shows the possibility of realizing FAST with a complete superconductor set of coils. An innovative active system is under development to reduce and to control the magnetic ripple. Tungsten (W) or liquid lithium (L–Li) has been chosen for the divertor material plates and the code EDGE2D has been used to optimize the divertor geometry.

Engineering evolution of the FAST machine

FAST (Fusion Advanced Studies Torus) 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 exploit innovative plasma facing component systems for DEMO. FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity δ = 0.4) and cost effective machine able to investigate, with integration capability, non linear dynamics effects of alpha particle behaviour in burning plasmas. FAST operates in high performance H-mode ( B T up to 8.5 T; I P up to 8 MA), as well as in advanced tokamak mode ( I P = 3 MA), and in full non inductive current mode ( I P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets. This allows for a pulse duration up to 170 s at 3 MA/3.5 T. The vacuum vessel (VV), segmented into 20-degree modules, is capable to accommodate a 40 MW RF power system. The machine has been designed to house a 10 MW Negative Neutral Beam Injection (NNBI) system. Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plate materials, and argon or neon as the impurities to be injected for mitigating the thermal loads.

Analytical modelling of resistive wall mode stabilization by rotation in toroidal tokamak plasmas

Stabilization of the resitive wall mode (RWM) may allow fusion power to be doubled for a given magnetic field in advanced tokamak operation. Experimental evidence from DIII-D and other machines suggests that plasma rotation can stabilize the RWM. Several authors (Finn 1995 Phys. Plasmas 2 3782, Bondeson and Xie 1997 Phys. Plasmas 4 2081) have constructed analytical cylindrical models for the RWM, but these do not deal with toroidal effects. The framework of Connor et al (1988 Phys. Fluids 31 577) is used to develop ideal plasma analytic models with toroidicity included. Stepped pressure profiles and careful ordering of terms are used to simplify the analysis. First, a current driven kink mode model is developed and a dispersion relation for arbitrary current profile is calculated. Second, the external pressure driven kink mode is similarly investigated as the most important RWM arises from this mode. Using this latter model it is found that the RWM is stabilized by Alfvén continuum damping with rotation levels similar to those seen in experiments. An expression for the stability of the external kink mode for more general current profiles and a resistive wall is derived in the appendix.

A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR) best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

Ideal MHD stability of a spherical tokamak power plant and a component test facility

We have investigated ideal MHD stability of two advanced spherical tokamak confinement concepts: the spherical tokamak power plant (STPP), a 3 GW concept fusion power plasma producing 1 GW of electric power, and the component test facility (CTF), a concept designed for in situ materials testing for ITER and beyond. Detailed stability studies for toroidal mode number n = 1, 2, 3 displacements are presented as a function of conformal wall radius Rw and on-axis safety factor q0. For the STPP marginal stability scans held the current profile fixed, but varied the total plasma current. For the CTF we have extended and parallelized earlier marginal stability scans to scan over both the plasma beta and q0 by varying the current profile to preserve the total plasma current. These confirm that both concepts are stable provided that the wall is sufficiently close and q0 sufficiently large (q0 > 2.8 for the power plant and q0 > 2.1 for the CTF). Both power plant and CTF configurations are found to be ballooning stable.

Analysis of high β regimes for DEMO

Stability limit calculations are presented for a range of tokamak power plant equilibria. The current drive requirements to sustain the optimised equilibrium profiles are confirmed by a transport code and the plasma shape is obtained from free-boundary equilibrium calculations. A pressure pedestal is included according to empirical scaling and ballooning mode stability limits. A terative optimisation of the profiles is undertaken to improve the baseline profiles in order to achieve the highest possible plasma performance and most favourable magnetohydrodynamic stability within conservative assumptions in order to increase confidence in the availability and control of the plasma. This results in a fully noninductive baseline operating scenario for a tokamak power plant design which has a broad low-shear q-profile which is meant to complement previous advanced tokamak design studies.

Residual gas analysis for long-pulse, advanced tokamak operation

A shielded residual gas analyzer (RGA) system on Tore Supra can function during plasma operation and is set up to monitor the composition of the neutral gas in one of the pumping ducts of the toroidal pumped limited. This "diagnostic RGA" has been used in long-pulse (up to 6 min) discharges for continuous monitoring of up to 15 masses simultaneously. Comparison of the RGA-measured evolution of the H(2)/D(2) isotopic ratio in the exhaust gas to that measured by an energetic neutral particle analyzer in the plasma core provides a way to monitor the evolution of particle balance. RGA monitoring of corrective H(2) injection to maintain proper minority heating is providing a database for improved ion cyclotron resonance heating, potentially with RGA-base feedback control. In very long pulses (>4 min) absence of significant changes in the RGA-monitored, hydrocarbon particle pressures is an indication of proper operation of the actively cooled, carbon-based plasma facing components. Also H(2) could increase due to thermodesorption of overheated plasma facing components.

Characterization of dusts and impurities from the interaction between the first wall materials and energetic laser beams

Study and control the formation mechanisms of dusts and impurities from the interaction between the edge plasma and the first wall materials are critical issues to achieve high performance in advanced tokamak (such as ITER). In this study, two major investigations have been carried out. Firstly, we have characterized the morphology and size of the dust particles formed from the interaction between the first wall materials and energetic laser beams. By choosing different laser powers we investigated the number and average size of dust particles as a function of deposited laser energy densities. The results indicate that the number and size of dust particles for tungsten and graphite targets both increase with laser power, but there are some differences in details between tungsten and graphite targets. This implies there are different mechanisms of dust generation for these two materials. Secondly, a high sensitivity molecular beam Time-of-Flight mass spectrometer (TOF-MS) has been developed to directly measure intermediate impurities from the interactions. This enables many intermediate impurities including transient clusters and ions are observed. The observed impurity ions are C n +, n = 4–32. The intensity distribution of the measured C n + indicates that larger carbon impurities are formed by complicated ways: probably including both syntheses by step growth and fragmentation from nano/micro flakes.

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.