Search

Systematic toroidal modeling of the plasma response to the n = 1–4 (n is the toroidal mode number) resonant magnetic perturbation (RMP) field is carried out in order to understand the plasma-shaping effect on controlling the type-I edge-localized modes (ELMs) in tokamak experiments. Considered are large variations of the plasma elongation and triangularity at a fixed edge safety factor q a, for limiter plasmas with both single-null (SN) and double-null (DN) divertor-like boundary shapes. Numerical results assuming conformal 3D RMP coils show that (i) the optimum coil phasing between the upper and lower rows for ELM control becomes increasingly sensitive to the plasma elongation with higher-n toroidal spectra, (ii) the optimum coil phasing is however essentially independent of the plasma triangularity for all n = 1–4 RMP fields, (iii) with the same coil current and the optimum coil phasing, high elongation generally favors ELM control but it may be more challenging for plasmas with intermediate elongation and with lower-n (n = 1–2) RMPs, and (iv) higher triangularity is generally always better for ELM control with all n = 1–4 fields for both DN and SN divertor-like plasma boundary shapes.

The characteristics of the divertor particle and heat fluxes are investigated during ELM bursts in ELMy H-mode plasmas with the lower single null (LSN) configuration in Korea Superconducting Tokamak Advanced Research (KSTAR). The particle and heat fluxes are evaluated from the electric probe measurements at the divertor region. It is found that the peak amplitude of the divertor flux during an ELM burst obtained near the outer strike point (OSP) decreases up to about 20% as the ELM frequency increases by a factor of ∼6.5 due to the ELM mitigation and the plasma shaping, which is similar to the trend of the amplitude versus the frequency of the ELM observed in other tokamaks. The ELMs are mitigated by using several methods as magnetic perturbation (MP) field, supersonic molecular beam injection (SMBI) and electron cyclotron heating (ECH) at the edge region. In addition, the particle flux, evaluated at the far scrape-off layer (SOL) region, is less than 1% of the divertor particle flux. In this work, results from the experimental investigations of particle and heat fluxes during ELM bursts from the electric probe measurements at the divertor and far SOL regions are presented.

The aim of this paper is to study the interplay between plasma shaping and magnet systems in superconducting tokamaks. In particular, we study the optimization of the layout of the superconducting toroidal field (TF) coil system as a function of plasma elongation (κ xp ) and aspect ratio (A). By excluding the use of in-vessel control coils, we proceed by maximizing plasma elongation as a function of the aspect ratio to keep plasma operation safe with respect to vertical displacement events. In this respect, design criteria coming from tokamak operational experience as well as other designs are used as guidance. The layout of the superconducting magnet systems is then optimized to meet the key requirements of reference TF on the plasma axis and maximum TF ripple at the plasma boundary, volt-seconds availability, and plasma shaping while satisfying the constraints of conductor temperature margins, peak stresses in the TF coil central vault, and hoop stress in the central solenoid.

The effect of plasma shaping on scrape-off layer (SOL) plasma turbulence is investigated through a rigorous validation exercise. Two- and three-dimensional simulations of the SOL plasma dynamics in three TCV limited discharges are carried out with the GBS code. These discharges realize an almost circular magnetic equilibrium, an elongated equilibrium, and an elongated equilibrium with negative triangularity. For the three plasma discharges, three simulations are performed, considering (i) a three-dimensional model with an explicit dependence on elongation, triangularity, and the inverse aspect ratio; (ii) a circular three-dimensional model in the infinite aspect ratio limit; and (iii) a two-dimensional model, which assumes a circular magnetic equilibrium in the infinite aspect ratio limit, cold ions, and interchange driven turbulence in the sheath limited regime. Ten validation observables common to simulations and experimental measurements from a reciprocating probe located at the TCV outer midplane are identified, and the agreement between experimental and numerical results relative to each observable is evaluated. The composite metric introduced by Ricci et al. [Phys. Plasmas 18, 032109 (2011)] is then used to assess the overall agreement between simulations and experimental measurements. It is found that the shaping model implemented in GBS improves the description of SOL plasma turbulence, taking into account the impact of elongation and triangularity.

The plasma shape, in particular the triangularity (δ), impacts on the pedestal stability. A scan of δ including a variation of heating power (P heat) and gas puff was performed to study the behaviour of edge localised modes (ELMs) and the pre-ELM pedestal stability for different plasma shapes. Generally, at higher δ the pedestal top electron density (n e) is enhanced and the ELM repetition frequency (f ELM) is reduced. For all δ, the pedestal top n e is already fully established to its pre-ELM value during the initial recovery phase of the n e pedestal, which takes place immediately after the ELM crash. The lowering of the f ELM with increasing δ is related to longer pedestal recovery phases, especially the last pre-ELM phase with clamped pedestal gradients (after the recovery phases of the n e and electron temperature (T e) pedestal) is extended. In all investigated discharge intervals, the pre-ELM pedestal profiles are in agreement with peeling-ballooning (PB) theory. Over the investigated range of δ, two well-separated f ELM bands are observed in several discharge intervals. Their occurrence is linked to the inter-ELM pedestal stability. In both kinds of ELM cycles the pedestal evolves similarly, however, the 'fast' ELM cycle occurs before the global plasma stored energy (W MHD) increases, which then provides a stabilising effect on the pedestal, extending the inter-ELM period in the case of the 'slow' ELM cycle. At the end of a 'fast' ELM cycle the n e profile is radially shifted inwards relative to the n e profile at the end of a 'slow' ELM cycle, leading to a reduced pressure gradient. The appearance of two f ELM bands suggests that the pedestal becomes more likely PB unstable in certain phases of the inter-ELM evolution. Such a behaviour is possible because the evolution of the global plasma is not rigidly coupled to the evolution of the pedestal structure on the timescales of an ELM cycle.

We present recent results on steady states of translational symmetric and axisymmetric ITER relevant plasmas with incompressible sheared flow in connection with a generalized Grad-Shafranov (GGS) equation and on their stability. On the basis of a vari- ety of ITER pertinent solutions of the GGS equation constructed it turns out that the flow affects noticeably the equilibrium quantities in qualitative agreement with the advanced confinement regimes phenomenology. Also, application of a sufficient condition for linear stability indicates that stabilization is mainly caused by the variation of the magnetic field in the direction perpendicular to the magnetic surfaces, related to the magnetic shear, depends on the plasma shaping and is sensitive to the up-down asymmetry. In certain cases the sheared flow has a weaker stabilizing effect enhanced by the equilibrium nonlinearity.

In this work, we report the development of a new gyrokinetic code for full-f simulation of electrostatic turbulence in general tokamak geometry. Backward semi-Lagrangian scheme is employed for noise-free simulation of the gyrokinetic Vlasov equation with finite Larmor radius effects. Grid systems and numerical interpolations are implemented to deal with arbitrary equilibrium information given in a GEQDSK format file. In particular, we introduce an adaptive interpolation technique for fluctuating quantities, which have elongated structures along equilibrium magnetic fields. This field-aligned interpolation can reduce the required number of grid points to represent the fluctuating quantities. Also, it is shown that the new interpolation allows us to choose bigger time sizes with better simulation accuracy. Several benchmark simulation results are presented for comparison with previously known cases. It is demonstrated that the new code can reproduce the well known results of zonal flow and linear ITG instabilities in concentric circular equilibrium. It is also shown that the code can capture the effects of plasma shaping on the zonal flow and ITG instabilities, and the stabilization effects of the shaping result in significant up-shifts of the threshold of ion temperature gradient for ITG in nonlinear simulation.

The low aspect ratio spherical tokamak (ST) MEDUSA-CR is currently being re-commissioned at Instituto Tecnologico de Costa Rica. One of the proposed first tasks is to simulate the magnetic confinement considering the particular coils arrangement of the device. This work presents the resulting shape of the plasma cross-section. MEDUSA-CR main specifications are: plasma major radius R0 approx. 0.14 m, plasma minor radius a approx. 0.10 m, toroidal field at the vessel geometrical center BT smaller than 0.5 T, plasma current Ip lesser than 40 kA, central electron temperature Te lesser than 140 eV, discharge duration lesser than 3 ms [1]. The free boundary equilibrium solver Fiesta has been used to obtain the plasma shaping and some equilibrium parameters. The code works the magnetic confinement for the static Magneto-Hydro-Dynamic (MHD) equilibrium case solving the Grad-Shafranov equation. A central solenoid was added to the original magnetic configuration aiming to create a limiter with an ergodic behavior in order to enhance the plasma confinement [2], [3]. The resulting cross-section (without divertor, i.e., natural divertor) for the plasma volume is a “bean” shape. The elongation is found to be 1.43 and the triangularity 0.547. A bean shape cross section is found to produce a higher triangularity, which have a significant effect on MHD stability [4]. Further simulations with FIESTA could be used as proof of principle for the effect of the plasma shape on transport and stability.

The free-streaming plus recycling model (FSRM) has recently been developed to understand and predict tungsten gross erosion rates from the divertor during edge localized modes (ELMs). In this work, the FSRM was tested against experimental measurements of W sputtering during ELMs, conducted via fast WI spectroscopy. Good agreement is observed using a variety of controlling techniques, including gas puffing, neutral beam heating, and plasma shaping to modify the pedestal stability boundary and thus the ELM behavior. ELM mitigation by pellet pacing was observed to strongly reduce W sputtering by flushing C impurities from the pedestal and reducing the divertor target electron temperature. No reduction of W sputtering was observed during the application of resonant magnetic perturbations (RMPs), in contrast to the prediction of the FSRM. Potential sources of this discrepancy are discussed. Finally, the framework of the FSRM is utilized to predict intra-ELM W sputtering rates in ITER. It is concluded that W erosion during ELMs in ITER will be caused mainly by free-streaming fuel ions, but free-streaming seeded impurities (N or Ne) may increase the erosion rate significantly if present in the pedestal at even the 1% level. Impurity recycling is not expected to cause significant W erosion in ITER due to the very low target electron temperature.

An explanation of the observed improvement in H-mode pedestal characteristics with increasing core plasma pressure or poloidal beta, as observed in MAST and JET, is sought in terms of the impact of the Shafranov shift, d', on ideal ballooning MHD stability., Comment: 18 pages, 5 figures. This is an author-created, un-copyedited version of an article submitted for publication in Plasma Physics and Controlled Fusion. IoP Publishing Ltd is not responsible for any errors or omissions in this version of the manuscript or any version derived from it

This paper briefly reviews a compact toroid reactor concept that addresses critical issues for forming, stabilizing and sustaining a field reversed configuration (FRC) with the use of plasma merging, plasma shaping, conducting shells, neutral beam injection (NBI). In this concept, an FRC plasma is generated by the merging of counter-helicity spheromaks produced by inductive discharges and sustained by the use of neutral beam injection (NBI). Plasma shaping, conducting shells, and the NBI would provide stabilization to global MHD modes. Although a specific FRC reactor design is outside the scope of the present paper, an example of a promising FRC reactor program is summarized based on the previously developed SPIRIT (Self-organized Plasmas by Induction, Reconnection and Injection Techniques) concept in order to connect this concept to the recently achieved the High Performance FRC plasmas obtained by Tri Alpha Energy [Binderbauer et al, Phys. Plasmas 22,056110, (2015)]. This paper includes a brief summary of the previous concept paper by M. Yamada et al, Plasma Fusion Res. 2, 004 (2007) and the recent experimental results from MRX.

The MAST-U (Mega-Amp Spherical Tokamak Upgrade) experiment is now being designed and will involve significant advances over the existing MAST machine, including increasing the toroidal field from 0.52 to 0.78 Tesla, the solenoid flux swing from 0.9 to 1.7 volt-seconds and the pulse length from 0.5 to 5 seconds (at 0.54 Tesla toroidal field). To meet these and other enhancements, the existing MAST centre column sub-assembly will be completely replaced in MAST-U. This sub-assembly comprises (a) the centre rod which forms the inner legs of the TF circuit, (b) the solenoid which fits closely over the centre rod, (c) the “air side” coils which are for plasma shaping – one within the bore of the centre rod (Pc) and one outside the solenoid (Px), (d) the two stainless steel crowns which are keyed and bonded to the ends of the centre rod for torsional support and (e) axial supports for the solenoid and airside coils. This paper describes the detailed design of the centre rod and the solenoid magnets and the means to manage the increased twisting torque on the centre rod (138 kNm in MAST-U compared with 58 kNm on MAST) due to the interaction of these two magnets.

Analytical theories of the geodesic acoustic mode (GAM) are reviewed in the small-and large-orbit drift width limits, respectively. Different physics pictures in these two limits are displayed. As an example, these two analytical methods are employed to investigate the plasma shaping effect on the frequency and collisionless damping rate of the GAM.

This study explores the effects of plasma shaping on magnetohydrodynamic mode stability and rotational stabilization in a tokamak, including both plasma and wall resistivity. Depending upon the plasma shape, safety factor, and distance from the wall, the β-limit for rotational stabilization is given by either the resistive-plasma ideal-wall (tearing mode) limit or the ideal-plasma resistive-wall (resistive wall mode) limit. In order to explore this broad parameter space, a sharp-boundary model is developed with a realistic geometry, resonant tearing surfaces, and a resistive wall. The β-limit achievable in the presence of stabilization by rigid plasma rotation, or by an equivalent feedback control with imaginary normal-field gain, is shown to peak at specific values of elongation and triangularity. It is shown that the optimal shaping with rotation typically coincides with transitions between tearing-dominated and wall-dominated mode behavior.

Novel stellarator configurations have been developed for ARIES-CS. These configurations are optimized to provide good plasma confinement and flux surface integrity at high beta. Modular coils have been designed for them in which the space needed for the breeding blanket and radiation shielding was specifically targeted such that reactors generating GW electrical powers would require only moderate major radii (

The collisionless residual zonal flow level is calculated for toroidally axisymmetric plasmas with arbitrary aspect ratio and large elongation. For small aspect ratio, this level deviates largely from that predicted by the Rosenbluth–Hinton formula and retains a finite value. Plasma shaping such as elongation is favorable for retaining the residual zonal flow as well. The trapped particles make a major contribution to the neoclassical polarization.

The JT-60SA (Super Advanced) project is a joint project of the ITER Satellite Tokamak program and the National Centralized Tokamak program in Japan with missions of supporting ITER, complementing ITER and exploring advanced issues toward DEMO. JT-60SA is a tokamak with superconducting coils, equipped with a poloidal field coil system with wide plasma shape controllability, upper and lower divertors with different shapes, NBI and ECRF with heating power 41 MW and various heating methods, in-vessel coils for suppressing MHD instabilities. With these functions, possibilities of producing ELMy H-mode with improved confinement, full non-inductive current drive of high beta plasmas (β N =3. 7 at Ip=3.5 MA, β N =4.4 at Ip=2.4 MA) and break-even class plasmas necessary for accomplishing the mission have been confirmed. The engineering design of JT-60SA is being done taking large annual neutron production into account. Double skin walls filled with borated water or boron doped concrete are employed for the vacuum vessel and cryostat, respectively, for neutron shield. Divertors structures and first walls are being designed so as to be changed with remote handling systems in the high radiation circumference.

The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (βN), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high nGW (because of an increase in Ip), which allows efficient use of the capacity of high βN. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

The impact of plasma shaping on electron heat transport is investigated in TCV L-mode plasmas. The study is motivated by the observation of an increase in the energy confinement time with decreasing plasma triangularity which may not be explained by a change in the temperature gradient induced by changes in the geometry of the flux surfaces. The plasma triangularity is varied over a wide range, from positive to negative values, and various plasmas conditions are explored by changing the total electron cyclotron (EC) heating power and the plasma density. The mid-radius electron heat diffusivity is shown to significantly decrease with decreasing triangularity and, for similar plasma conditions, only half of the EC power is required at a triangularity of −0.4 compared with +0.4 to obtain the same temperature profile. Besides, the observed dependence of the electron heat diffusivity on the electron temperature, electron density and effective charge can be grouped in a unique dependence on the plasma effective collisionality. In summary, the electron heat transport level exhibits a continuous decrease with decreasing triangularity and increasing collisionality. Local gyro-fluid and global gyro-kinetic simulations predict that trapped electron modes are the most unstable modes in these EC heated plasmas with an effective collisionality ranging from 0.2 to 1. The modes stability dependence on the plasma triangularity is investigated.

The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which could damage the inner wall of the machine, while operating the device with larger margins against the operational limits would lead to a reduction of the electrical output power. Second, the performance of DEMO control is limited by space restrictions for the implementation of components (optimization of the tritium breeding rate), by lifetime issues for the front-end parts (neutron and gamma radiation, erosion and deposition acting on all components) and by slow, weak and indirect action of the available actuators (plasma shaping, heating and fuelling). The European DEMO conceptual design studies include the development of a reliable control system, since the details of the achievable plasma scenario and the machine design may depend on the actual performance of the control system. In the first phase of development, an initial understanding of the prime choices of diagnostic methods applicable to DEMO, implementation and performance issues, the interrelation with the plasma scenario definition, and the planning of necessary future R&D have been obtained.

The EDT2 750V uses a micro pattern trench cell with a narrow mesa for reducing the on-state losses with a tailored channel width for short circuit robustness. To account for high system stray inductances (L stray ) and currents for Full or Hybrid Electric Vehicle inverter applications, it features a 750V voltage rating compared to the predecessor IGBT3 650V by an optimized vertical structure and proper plasma shaping. This plasma distribution not only determines the performance tradeoff between on-state and switching losses, but at the same time defines the surge voltage for a given L stray ∗I in the application as visualized in a switch-off loss vs. surge voltage trade-off diagram. Shaping of the feedback capacitance Cgc optimizes the tunability of the switching slopes by means of an external gate resistor for an easier adaption to a wider range of system inductances with low losses.

Here the recent developments in GEM, a quite comprehensive Gyrokinetic Electromagnetic (GEM) turbulence code are described. GEM is a δf particle turbulence simulation code that has kinetic electrons and electromagnetic perturbations. The key elements of the GEM algorithm are: (1) the parallel canonical formulation of the gyrokinetic system of equations (1); (2) an adjustable split-weight scheme for kinetic electrons (2); and (3) a high-β algorithm for the Ampere's equation (3). Additionally, GEM use a two-dimensional (2D) domain decomposition and runs efficiently on a variety of high performance architectures. GEM is now extended to include arbitrary toroidal equilibrium profiles and flux-surface shapes (4). The domain is an arbitrarily sized toroidal slice with periodicity assumed in the toroidal direction. It is global radially and poloidally along the magnetic field. Results are presented that demonstrate the effect of plasma shaping on the Ion-Temperature-Gradient (ITG) driven instabilities. An example of nonlinear simulation of the finite-β modified ITG turbulence in general geometry is also given. Finally, collisionless Trapped Electron Modes (TEM) are investigated and shown here is the transition from the TEM dominated core region to the Drift-Wave dominated edge region as the density gradient increases.

Systematic and statistical studies have been conducted in order to develop an understanding of the parametric dependences of both the global and thermal energy confinement times at low aspect ratio in high power National Spherical Torus Experiment discharges. The global and thermal confinement times of both L- and H-mode discharges can exceed values given by H-mode scalings developed for conventional aspect ratio. Results of systematic scans in the H-mode indicate that the confinement times exhibit a nearly linear dependence on plasma current and a power degradation weaker than that observed at conventional aspect ratio. In addition, the dependence on the toroidal magnetic field is stronger than that seen in conventional aspect ratio tokamaks. This latter trend is also evident in statistical analyses of the available dataset. These statistical studies also indicate a weaker parametric dependence on plasma current than found in the systematic scans, due to correlations among the predictor variables. Regressions based on engineering variables, when transformed to dimensionless physics variables, indicate that the dependence of BτE on βt can range from being negative to null. Regressions based directly on the dimensionless physics variables are inexact because of large correlations among these variables. Scatter in the confinement data, at otherwise fixed operating parameters, is found to be due to variations in ELM activity, low frequency density fluctuations and plasma shaping.

This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter S(=q95*Ip /a*Bt) and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of ~ 7 was obtained by double null divertor configuration with a lower aspect ratio of A ~ 2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating

The ITER CS contains 6 identical but electrically independent modules. ITER plasma shaping requirements yield a wide range of radial (or hoop) loads and a wide range of vertical loads on each CS module. Pre-compression is used to maintain contact between the modules during all stages of plasma operation, to minimize primary tensile stresses in the coil insulation, and to prevent toroidal ratcheting of the modules during repeat operation. The preload requirements for the ITER Central Solenoid (CS) are the subject of this paper. Axial pre-compression of CS stack is carefully analyzed using finite element (FE) models. Conflicting requirements on the pre-compression structure (namely, the generation of large axial pre-compression with minimal use of radial space outside the modules) are balanced to support the desired plasma shapes

Fusion DEMO plant design at JAERI directs toward compactness and economy of the reactor with a fusion output at 3 GW level. The smaller size of the center solenoid (CS) has a large impact on reducing the reactor size and weight. Three candidates for CS, namely “-less”, “slim” and “full” CS, are considered according to the role of CS coil that is plasma ramp up and plasma shaping. From points of views of technical feasibility, obtained triangularity (δx) and weight of the reactor, the slim-CS is selected as the candidate for DEMO at JAERI. For slim-CS, high δx (≥0.43), which is enough to suppress giant ELMs, is obtained, and highly radiative outer divertor is required in steady state power handling. The consistent equilibrium was found to satisfy the divertor requirements and radial build. During the plasma current ramp up phase, the acceleration voltage of neutral beam injection (NBI) has to be changed more than two times. In order to inject sufficient power during the low-energy beam injection, the present technology must be replaced by a new constant-current, variable-voltage technology.

Three options of fusion DEMO plant are proposed characterized by functions of the center solenoid (CS). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3 GW with a major radius of 5.5 m, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4 T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb3Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper.

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6 - approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βtwith nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented. © 2006 IAEA, Vienna.

A key feature of the 'high recycling steady' (HRS) H-mode regimes is the reduction of the transient heat load to the divertor target due to large ELMs. The heat flux to the divertor targets measured by means of a Langmuir probe in the scrape-off layer on the target plate is about ~0.3 MW m−2 in the typical HRS plasmas, which is very small in comparison with the transient heat load in the ELMy H-mode discharge. In addition, the effects of the plasma shaping on the access to and improved plasma performance of HRS H-mode regimes have been investigated. Furthermore, simultaneous measurements of both density fluctuations and potential profiles at the plasma edge region using a heavy ion beam probe are also presented.

Synopsis: The KSTAR project was started in 1996 and conducted in three phases, the conceptual design phase (1996–1998), the R&D and engineering design phase (1998–2002), and the construction phase (2002–2007). The mission of the Korea Superconducting Tokamak Advanced Research (KSTAR) Project is to develop a steady-state-capable advanced superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Since the KSTAR mission includes the achievement of a steady-state-capable operation, the use of superconducting coils is an obvious choice for the magnet system and the long pulse current drive and heating systems are also the important aspect of the KSTAR design features. The advanced tokamak aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizer, internal control coils. All the major components are in the stage of the fabrication and assembly for the completion of the KSTAR construction in the year 2007.

Integrated scenario simulations are done for NSTX that address four primary objectives for developing advanced spherical torus (ST) configurations: high ? and high ?N inductive discharges to study all aspects of ST physics in the high ? regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current drive techniques; non-inductively sustained discharges at high ? for flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX and non-solenoidal startup and plasma current rampup. The simulations done here use the tokamak simulation code and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral beam deposition profile and other characteristics. CURRAY is used to calculate the high harmonic fast wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal MHD stability is done with JSOLVER, BALMSC and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with ?T ? 40% at ?N's of 7.7?9, IP = 1.0?MA and BT = 0.35?T. The plasma is 100% non-inductive and has a flattop of four skin times. The resulting global energy confinement corresponds to a multiplier of H98(y),2 = 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control and early heating/H-mode transition for producing and optimizing these plasma configurations.

The overall vision for FIRE is to develop and test the fusion plasma physics and plasma technologies needed to realize capabilities of the ARIES-RS/AT power plant designs. The mission of FIRE is to attain, explore, understand and optimize a fusion dominated plasma which would be satisfied by producing DT fusion plasmas with nominal fusion gains {approx}10, self-driven currents of {approx}80%, fusion power {approx} 150 - 300 MW and pulse lengths up to 40 s. Achieving these goals will require the deployment of several key fusion technologies under conditions approaching those of ARIES-RS/AT. The FIRE plasma configuration with strong plasma shaping, a double null pumped divertor and all metal plasma facing components is a 40% scale model of the ARIES-RS/AT plasma configuration. 'Steady-state' advanced tokamak modes in FIRE with high {beta}, high bootstrap fraction and 100% non-inductive current drive are suitable for testing the physics of the ARIES-RS/AT operating modes. The development of techniques to handle power plant relevant exhaust power while maintaining low tritium inventory is a major objective for a burning plasma experiment. The FIRE H-modes and AT-modes result in fusion power densities from 3 - 10 MWm{sup -3} and neutron wall loading from 2 - 4 MW m{supmore » -2} which are at the levels expected from the ARIES-RS/AT design studies.« less

A variety of techniques are being tested on ASDEX Upgrade for avoiding or reducing the divertor energy load caused by ELMs. Quiescent H-mode operation has been improved by operation with recently boronized vessel walls. Zeff values in QH-modes are now ~2.5, comparable with those found in ELMy H-modes of similar density. Type II ELM operation has been generated in discharges with ICRF additional heating alone, thus demonstrating its compatibility with reactor-relevant low core momentum and particle input. Furthermore, Type II ELM operation at low (

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with ?N ? 3.5 and ?T ? 3.6% sustained for up to 1?s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by results of these and other recent experiment and simulation efforts.

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these condi...

In ASDEX Upgrade, experimental efforts aim to establish pace making and mitigation of type-I edge localized modes (ELMs) in high confinement mode (H-mode) discharges. Injection of small size cryogenic deuterium pellets (~(1.4?mm)2 ? 0.2?mm ? 2.5 ? 1019?D) at rates up to 83?Hz imposed persisting ELM control without significant fuelling, enabling for investigations well inside the type-I ELM regime. The approach turned out to meet all required operational features. ELM pace making was realized with the driving frequency ranging from 1 to 2.8 times the intrinsic ELM frequency, the upper boundary set by hardware limits. ELM frequency enhancement by pellet pace making causes much less confinement reduction than by engineering means like heating, gas bleeding or plasma shaping. Confinement reduction is observed in contrast to the typical for engineering parameters. Matched discharges showed triggered ELMs ameliorated with respect to intrinsic counterparts while their frequency was increased. No significant differences were found in the ELM dynamics with the available spatial and temporal resolution. By breaking the close correlation of ELM frequency and plasma parameters, pace making allows the establishment of fELM as a free parameter giving enhanced operational headroom for tailoring H-mode scenarios with acceptable ELMs. Use was made of the pellet pace making tool in several successful applications in different scenarios. It seems that further reduction of the pellet mass could be possible, eventually resulting in less confinement reduction as well.

Results from recent experiments to study the effects of divertor geometry and increased plasma shaping on the L–H transition threshold on JET are reported. Equivalent septum configurations run with the new septum replacement plate (SRP) in the MkII Gas Box divertor have shown that the presence of the septum lowers the L–H transition power threshold, Pth, by 20%. For X-point to virtual septum top distances of less than 6 cm, the SRP plasmas also demonstrate a significant decrease of the L–H Pth and pedestal electron temperature, Te with reduced X-point height. Although, the SRP plasma's Pth remains above that with the septum, there is no difference in the pedestal Te at the L–H transition. The influence of plasma shaping on the L–H transition has also been investigated for the first time on JET in a series of density scans at Ip/Bt of 2.5 MA/2.7 T. While keeping the lower triangularity, δlower, and divertor geometry constant, the upper triangularity, δupper, has been increased from 0.23 to 0.34 with no effect on Pth or the pedestal Te or Ti. In a separate edge ne scan, two configurations with different δupper/δlower values of 0.23/0.23 and 0.43/0.33 have been compared. A large difference in the transition threshold is observed at values of edge ne above 1.8 × 1019 m−3, with the higher δ plasmas characterized by Pth of up to 25% lower at the highest densities scanned. This is thought to be the result of lowered X-point and outer strike point heights with increased δ.

Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks, leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments (Scoville J.T. and La Haye R.J. 2003 Nucl. Fusion 43 250) and resistive wall mode experiments (Garofalo A.M., La Haye R.J. and Scoville J.T. 2002 Nucl. Fusion 42 1335) done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made in DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported (La Haye R.J. and Scoville J.T. 1991 Rev. Sci. Instrum. 61 2146). Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the Ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all the coils are one-third the values indicated from the stability experiments (Garofalo et al 2002, Scoville and La Haye 2003). These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.

In this paper, we demonstrate the importance of the details of the equilibria on the stability of electron drift waves. A comparison of electrostatic electron drift waves in numerical and analytical tokamak equilibria is presented in fully three-dimensional circular and non-circular tokamaks. The numerical equilibria are obtained using the variational moments equilibrium code and the analytical equilibria used is the generalized ŝ-α model. An eigenvalue equation for the model is derived using the ballooning mode formalism and solved numerically using a standard shooting technique. The stability and the localization of the electron drift wave is found to be strongly dependent on the local shear of the magnetic field. Large values of the local shear are found to be stabilizing. A disagreement in the results is found between analytical and numerical equilibria at aspect ratios of typical tokamaks, suggesting that the latter approach should be used in the transport calculations. The effects of the local shaping of the magnetic surfaces are complicated and can be both stabilizing and destabilizing, depending on the details of the equilibria.

Non-axisymmetric (error) fields in tokamaks lead to a number of instabilities including so-called locked modes [J.T. Scoville, R.J. La Haye, Nucl. Fusion 43 (4) (2003) 250–257] and resistive wall modes (RWM) [A.M. Garofab, R.J. La Haye, J.T. Scoville, Nucl. Fusion 42 (11) (2002) 1335–1339] and subsequent loss of confinement. They can also cause errors in magnetic measurements made by point probes near the plasma edge, error in measurements made by magnetic field sensitive diagnostics, and they violate the assumption of axisymmetry in the analysis of data. Most notably, the sources of these error fields include shifts and tilts in the coil positions from ideal, coil leads, and nearby ferromagnetic materials excited by the coils. New measurements have been made of the n=1 coil-related field errors in the DIII–D plasma chamber. These measurements indicate that the errors due to the plasma shaping coil system are smaller than previously reported and no additional sources of anomalous fields were identified. Thus they fail to support the suggestion of an additional significant error field suggested by locked mode and RWM experiments.

Transport and plasma chemistry in magnetized plasma sources for plasma processing are become increasingly important. Beyond the plasma shaping and confinement associated with magnetic fields in magnetically enhanced reactive ion etching (MERIE) tools, magnetized inductively coupled plasmas and conventional magnetrons, there are electron-kinetics-driven chemistry functions as well. It is difficult to capture these effects using conventional fluid or hybrid approaches, as it is not straightforward how to map transport and kinetic macroparameters onto common orthogonal grids. This paper addresses this issue by characterizing macroparameters (also known as swarm parameters) as a function of crossed electric and magnetic field crossing angle for Cl2 and SiH4. Scaling or mapping criteria (i.e., mapping of swarm parameters onto an orthogonal numerical grid) are described to be a function of the relative position of the energy threshold of collision processes and the mean electron energy. Using these criteria, the transport or kinetic parameters may be mapped either by angle cosines or by their components on an orthogonal grid. Significant exceptions to this occur for certain inelastic processes. When threshold energies coincide with the mean electron energy, it is not possible to determine simple mapping criteria independent of exact knowledge of how the parameters behave at a specific crossing angle.