High Toroidal Field Research Articles

High Magnetic Shear Gain in a Liquid Sodium Stable Couette Flow Experiment: A Prelude to anα−ΩDynamo

The Ω phase of the liquid sodium α-Ω dynamo experiment at New Mexico Institute of Mining and Technology in cooperation with Los Alamos National Laboratory has demonstrated a high toroidal field B(ϕ) that is ≃8×B(r), where B(r) is the radial component of an applied poloidal magnetic field. This enhanced toroidal field is produced by the rotational shear in stable Couette flow within liquid sodium at a magnetic Reynolds number Rm≃120. Small turbulence in stable Taylor-Couette flow is caused by Ekman flow at the end walls, which causes an estimated turbulence energy fraction of (δv/v)(2)∼10(-3).

The Fusion Advanced Studies Torus (FAST): a proposal for an ITER satellite facility in support of the development of fusion energy**This article was originally submitted as a paper relating to the 2008 Fusion Energy Conference (Geneva). The Cumulative Index of these papers can be found in issue 3 2010 of Nuclear Fusion http://iopscience.iop.org/0029-5515/50/3/039901.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (BT up to 8.5 T) and plasma current (Ip up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30 MW on the plasma). Moreover, the experiment foresees the use of 6 MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4 MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative neutral beams (NNBI) in the energy range 0.5–1 MeV. The total power input will be in the 30–40 MW range under different plasma scenarios with a wall power load comparable to that of ITER (P/R ∼ 22 MW m−1). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to ≈1.5), to a full current drive scenario, lasting around 170 s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant operation regimes. The divertor itself is designed to be completely removable by remote handling. This will allow us to study (in view of DEMO) the behaviour of innovative divertor concepts, such as those based on liquid lithium.FAST is capable of operating with very long pulses, up to 170 s, despite being a copper machine. The magnets initial operation temperature is 30 K, with cooling provided by helium gas. The in vessel components, namely FW and divertor, are actively cooled by pressurized water above 80 °C. The same water is also used to bake the vacuum vessel. FAST is equipped with ferromagnetic inserts to keep the toroidal field magnet ripple down to 0.3%.

Momentum transport studies in JET H-mode discharges with an enhanced toroidal field ripple

In this study, enhancement of the toroidal field (TF) ripple has been used as a tool in order to reveal the impact of the momentum pinch on the rotation profiles in H-mode JET discharges. The analysis showed that flatter rotation profiles were obtained in discharges with a high TF ripple, attributed to a smaller inward momentum convection. An average inward momentum pinch of approximately Vp ≈ 3.4 m s−1 and a normalized pinch value of RVp/χ ≈ 6.6 could explain the observation. The data show that the momentum at the edge affects the peaking of the rotation and momentum density profiles. Under the assumption that the heat and momentum diffusivities are equal, an estimate of the levels of the momentum pinch in all discharges in the JET rotation database was made. For H-mode discharge these ranged from 0.3 m s−1 < Vp < 17 m s−1, with 2 < RVp/χ < 10. A larger momentum pinch was found in discharges with a smaller density profile gradient length, i.e. a more peaked density profile.

Vacuum and wall conditioning system on EAST

EAST is a non-circular advanced steady-state experimental device and the first entirely superconducting tokamak in the world. Vacuum system is one of the most important sub-systems of EAST device. Wall conditionings, such as baking, discharge cleaning and boronization, also play a very important role for the plasma operation. Due to Ion Cyclotron Resonance Frequency (ICRF) wall conditioning technique could be carried out in the presence of a high toroidal magnetic field, it is routinely used for wall cleanings on EAST. After the 2nd campaign in 2007, the plasma facing walls was modified to full carbon walls and vacuum system was upgraded to meet the requirement of particles exhaust. This paper will introduce the new statues of vacuum system, such as pumping, fueling and wall conditionings on EAST. Then the vacuum operation and wall conditioning in the 2008 campaign is introduced.

The design, analysis and alignment of EAST divertor

The divertor of EAST is composed of inner-target, outer-target and dome. Every kind of target has 16 separate modules along toroidal direction. The structure of divertor includes support, heat sink and the first wall graphite. The shape and dimension of divertor are optimized for flexible operation of different elongation and triangularity of plasma. The eddy current and halo current interact with high toroidal magnetic field will produce large forces on divertor, so the structure of divertor must bear these electromagnetic forces. The first wall of divertor must have high assembly accuracy of 0.5 mm. The traditional method using optical instruments to survey directly is impractical because of the especial location of divertor. The benchmark templet integrates with optical instruments to align the divertor is proved simple and effective.

Vacuum operation of first divertor campaign on EAST

Experimental Advanced Superconducting Tokamak is the first whole superconducting tokamak with divertor configuration commissioned in the world. Vacuum system is one of the most important sub-systems of EAST device. Wall conditioning also plays a very important role for the plasma operation. Because Ion Cyclotron Resonance Frequency discharge cleaning could be carried out with a high toroidal magnetic field, it is used often for wall conditioning, especially for boronization, which was done successfully on EAST. This paper describes the vacuum operation of this new device. The first part describes the vacuum system, consisting of pumping, fuelling and wall conditioning related sub-systems. Then the operation of the vacuum system is introduced including leak detecting, pumping, boronization, etc. At last, the plan for the coming campaign is discussed.

Prospects for steady-state scenarios on JET

In the 2006 experimental campaign, progress has been made on JET to operate non-inductive scenarios at higher applied powers (31 MW) and density (nl ∼ 4 × 1019 m−3), with ITER-relevant safety factor (q95 ∼ 5) and plasma shaping, taking advantage of the new divertor capabilities. The extrapolation of the performance using transport modelling benchmarked on the experimental database indicates that the foreseen power upgrade (∼45 MW) will allow the development of non-inductive scenarios where the bootstrap current is maximized together with the fusion yield and not, as in present-day experiments, at its expense. The tools for the long-term JET programme are the new ITER-like ICRH antenna (∼15 MW), an upgrade of the NB power (35 MW/20 s or 17.5 MW/40 s), a new ITER-like first wall, a new pellet injector for edge localized mode control together with improved diagnostic and control capability. Operation with the new wall will set new constraints on non-inductive scenarios that are already addressed experimentally and in the modelling. The fusion performance and driven current that could be reached at high density and power have been estimated using either 0D or 1–1/2D validated transport models. In the high power case (45 MW), the calculations indicate the potential for the operational space of the non-inductive regime to be extended in terms of current (∼2.5 MA) and density (nl > 5 × 1019 m−3), with high βN (βN > 3.0) and a fraction of the bootstrap current within 60–70% at high toroidal field (∼3.5 T).

Discussion About the Size of a Future Fusion Demonstration Reactor: The Impact of the Toroidal Magnetic Field

In parallel to ITER preparation, power plant conceptual studies have been carried out in Europe as a guideline for the main design parameters of a fusion demonstration reactor, the construction of which is scheduled to start twenty years hence. A key component of this fusion demonstration reactor is the superconducting magnetic system, which represents in ITER 30% of the investment cost. A high toroidal confinement field Bt, can help to moderate the major radius of the tokamak, and thereby the investment cost of the machine. Two versions of such a demonstration reactor, with different values of Bt will be compared in terms of volume, radial extension for the toroidal field magnet inner leg, central solenoid available flux swing and investment cost.

Pellet ablation and mass deposition in FTU: analysis of vertical and low field side injection experiments

Pellet injection is considered the most promising technique to fuel ITER plasmas. However, standard low field side (LFS) pellet injection seems inadequate for an efficient fuelling of the plasma core because it is too demanding in terms of injection velocity. Alternative configurations, as injection from the high field side (HFS) or vertically from the top of the machine (VHFS) are technologically less demanding due to a drift of the ablated material down the magnetic field gradient, which helps its penetration. In the past year LFS and VHFS pellet injection experiments have been performed on the Frascati Tokamak Upgrade (FTU: BT = 8 T, Ip = 1.5 MA, R = 93.5 cm, a = 30 cm), a circular cross-section device with a molybdenum toroidal limiter located on the HFS. The rationale of these experiments was to study the advantages of VHFS injection in a high toroidal field machine, where the value of BT can be considered as representative of that in ITER. The results demonstrate that the main advantage of VHFS injection (with respect to LFS) is that it allows one to obtain the same post-pellet density profile with an injection speed 2–3 times lower than that required in the case of LFS injection. However no significant displacement of the ablated material was observed, possibly due to the small value of the ∇ B-induced drift in these ohmic, low β plasmas. The experimental measurements were compared to the predictions of the drift model (Pégourié B. et al 2007 Nucl. Fusion 47 44) showing a reasonable agreement in both the magnitude of the drift and the main features of the line integrated density during pellet ablution.

Tokamak-like confinement at a high beta and low toroidal field in the MST reversed field pinch**Presented at the 19th IAEA Fusion Energy Conference with the title Overview of Improved Confinement and Plasma Control in the MST Reversed Field Pinch.

Energy confinement comparable with tokamak quality is achieved in the Madison Symmetric Torus (MST) reversed field pinch (RFP) at a high beta and low toroidal magnetic field. Magnetic fluctuations normally present in the RFP are reduced via parallel current drive in the outer region of the plasma. In response, the electron temperature nearly triples and beta doubles. The confinement time increases ten-fold (to ∼10 ms), which is comparable with L- and H-mode scaling values for a tokamak with the same plasma current, density, heating power, size and shape. Runaway electron confinement is evidenced by a 100-fold increase in hard x-ray bremsstrahlung. Fokker–Planck modelling of the x-ray energy spectrum reveals that the high energy electron diffusion is independent of the parallel velocity, uncharacteristic of magnetic transport and more like that for electrostatic turbulence. The high core electron temperature correlates strongly with a broadband reduction of resonant modes at mid-radius where the stochasticity is normally most intense. To extend profile control and add auxiliary heating, rf current drive and neutral beam heating are in development. Low power lower-hybrid and electron Bernstein wave injection experiments are underway. Dc current sustainment via ac helicity injection (sinusoidal inductive loop voltages) is also being tested. Low power neutral beam injection shows that fast ions are well-confined, even in the presence of relatively large magnetic fluctuations.

H-mode research in NSTX

H-modes are routinely obtained in the National Spherical Torus Experiment (NSTX) and have become a standard operational scenario. L–H transitions triggered by NBI heating have been obtained over a wide parameter range in Ip, Bt, and n̄e in either lower-single-null (LSN) or double-null (DN) diverted discharges. Edge localized modes are observed in both configurations but the characteristics differ between DN and LSN, which also have different triangularities (δ). An H-mode duration of 500 ms was obtained in LSN, with a total pulse length of ∼1 s. Preliminary power threshold studies indicate that the L–H threshold is between 600 kW and 1.2 MW, depending on the target parameters. Gas injector fuelling from the centre stack (i.e. the high toroidal field side) has enabled routine H-mode access, and comparisons with low-field side (LFS) fuelled H-mode discharges show that the LFS fuelling delays the L–H transition and alters the pre-transition plasma profiles. Gas puff imaging and reflectometry show that the H-mode edge is usually more quiescent than the L-mode edge. Divertor infrared camera measurements indicate up to 70% of available power flows to the divertor targets in quiescent H-mode discharges.

Research Activities on Tokamaks in Japan: JT-60U, JFT-2M, and TRIAM-1M

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state operation. Plasma performance in transient and quasi steady states has been significantly improved, utilizing reversed shear and weak shear (high-βp) ELMy H-modes characterized by both internal and edge transport barriers and high bootstrap current fractions. Development of each key issue for advanced steady-state operation has also been advanced. Advanced and basic research of JFT-2M has been performed to develop high-performance tokamak plasma as well as the structural material for a fusion reactor. Toroidal field ripple reduction with ferritic steel plates outside the vacuum vessel is successfully demonstrated. No adverse effects to the plasma were observed with poloidal fields inside the vacuum vessel (partial covering). Preparation is in progress for full-scale testing of the compatibility of the ferritic steel wall (full covering) with plasma. A heavy ion beam probe has been installed to study H-mode plasmas. Compact toroid (CT) injection experiments are performed, showing deep CT penetration into the core region of the H-mode. The TRIAM project has investigated steady-state operation and high-performance plasma of a tokamak with the high toroidal magnetic field superconducting tokamak. Four important contributions in the fields of fusion technology of superconducting tokamaks, steady-state operation, high-performance plasma, and startup of plasma current without the assistance of center solenoid coils have been achieved on TRIAM-1M, especially regarding steady-state operation by realization of a discharge for >3 h.

β limiting MHD activity and mode locking in Alcator C-Mod

Since Alcator C-Mod normally operates at high toroidal field and high collisionality, β limiting instabilities are rarely observed. Under some conditions, however, when operating at low collisionality and high input power (PICRF⩽5 MW), large amplitude (5×10-5<[B̃θ/Bθ]wall⩽5×10-3) low-frequency (fMHD<50 kHz) MHD modes appear to limit the achievable β. Modes with m/n = 5/4,4/3,3/2, and 2/1 were destabilized when βp>0.52 and increased in amplitude with increasing β until a rollover or collapse in β occurred. The largest amplitude modes with m = 2,n = 1 strongly degraded momentum and energy confinement when the modes coupled across the plasma core and locked to the wall, bringing the plasma ion toroidal rotation to zero, within experimental errors, about 50 ms after mode locking. MHD stability was calculated with the linear resistive toroidal MHD code MARS for a discharge with a large m = 2,n = 1 mode. Comparisons with neoclassical tearing mode (NTM) theory and with NTMs found on other tokamaks indicate that these modes may be driven by a combination of resistive, neoclassical, and error field effects.

Equilibrium and stability properties of self-organized electron spiral toroids

A cold-fluid model for a self-organized electron spiral toroid (EST) is presented. In the present model, the electrons are assumed to undergo energetic spiral motion along a hollow torus with a fixed ion background, the electron mean free path is assumed to be long compared with the torus size, and the minor radius of the EST is assumed to be small compared with the major radius. Using this model, the equilibrium and stability properties of the electron flow in the self-organized EST are analyzed. It is found that a class of self-organized EST equilibria exists with or without an externally applied toroidal magnetic field. It is shown that in the absence of any applied toroidal magnetic field, the EST equilibria are stable at high electron densities (i.e., at high toroidal self-magnetic fields), although they are unstable at low electron densities (i.e., at low toroidal self-magnetic fields).

RF wall conditioning – a new technique for future large superconducting tokamak

A new technique for wall conditioning with a high toroidal magnetic field has been developed on HT-7 superconducting tokamak by launching ion cyclotron resonant wave (ICRF) into plasmas. The RF wall conditioning techniques include hydrogen removal, isotopic and impurity control, boronization and siliconization, which have been proved to be very effective and could be routinely used in daily experiments.

Measurements of Escaping Fast Particles Using a Thin-Foil Charge Collector

Two screened, thin-foil charge collectors were mounted just beyond the plasma edge at an outboard position (below midplane) in the Joint European Torus to detect lost alpha particles during the 1997 high fusion power D-T experiments. No convincing observations of alpha-particle collection were obtained, possibly because of the low level of alpha-particle losses but more probably because the positioning of the detector was not ideal for the high fusion power discharges that were run at high plasma current and toroidal field. Under such conditions, alpha particles on escaping orbits leading toward the detector are highly likely to be intercepted by the nearby poloidal limiter. Moreover, a small alpha-particle signal would have been obscured by interference from a large and unexpected signal attributed here to fast neutrals, leaving the plasma and ionizing in the low density scrape-off region outside the plasma boundary. The interpretation of this unexpected signal is discussed. In all probability, it will also be encountered in any future attempts to detect lost alpha particles in a current measuring detector unless suitable precautions are taken, e.g., provision of a thin first foil to remove light charged particles with energies below ~0.5 MeV.

Toroidal field coil design for the spherical tokamak power plant

The design of the toroidal field (TF) coil system of a spherical tokamak (ST) power plant has considerable influence on the design and performance of the entire plant due mainly to its electrical power requirements which are dominated by the centre rod conductor. The low aspect ratio and high TF coil current combined with the effects of neutron irradiation on the structures lead to compromises in the centre column dimensions and choice of materials. Since the electrical conductivity of metals increases markedly with falling temperatures this offers a route for reducing the resistive power dissipation in the coils. However the coils now need refrigerating which also consumes power which adds to the total dissipation. The optimum for a water cooled copper centre conductor appears to be around room temperature and the optimum for a helium cooled aluminium conductor appears to be around 30 K. Centre column and TF coil designs based on these two options are being developed and the current status of the designs is presented here. The water cooled copper option uses a dispersion strengthened copper alloy for the centre rod with minimal shielding giving a simple design. The cryogenically cooled option uses pure aluminium and offers advantages in terms of power consumption but requires significant shielding and more complex structural supports.

Joint Instability of Latitudinal Differential Rotation and Toroidal Magnetic Fields below the Solar Convection Zone. II. Instability for Toroidal Fields that Have a Node between the Equator and Pole

We generalize results of Gilman and Fox to unperturbed toroidal fields that have a node somewhere between the equator and the pole as we speculate the Sun's field to have for most phases of its magnetic cycle. We use the same solution method as in Gilman and Fox, namely Legendre polynomial expansion and matrix inversion to solve for the eigenvalues and eigenfunctions. The solutions are structured around certain singular or critical points of the suitably transformed and combined vorticity and induction equations. There are singular points at the poles, and singularities where ω0-cr=±α0, in which ω0 is the local rotation rate, cr is the longitudinal phase speed of an unstable wave, and α0 is an angular measure of the toroidal field. We survey the instability as a function of toroidal field profile and amplitude as well as differential rotation amplitude, thereby examining reference states that could be characteristic of most phases of the solar cycle, and most depths within the rotational shear layer just below the base of the solar convection zone. As found in Gilman and Fox, instability occurs for a wide range of both toroidal fields and differential rotations. Differential rotation is again the primary energy source for growing modes when the toroidal field is weak, and the toroidal field is the primary source when it is strong. Unlike in Gilman and Fox, here modes of both symmetries about the equator are unstable for low and high toroidal fields, and for high fields a second antisymmetric mode appears. Which mode symmetry is favored for low fields depends in detail on the relative amplitudes of differential rotation and toroidal field. For low toroidal fields (unstable) modes of both symmetries are energetically active (extracting energy from the unperturbed state) only poleward of the node and an adjacent singularity, but are coupled to energetically neutral velocity perturbations equatorward of that singular point. In transition to higher field strengths, those velocity patterns are damped out when two additional singular points appear in the system, but the energetically active high-latitude disturbances remain. By contrast the second antisymmetric mode is energetically active equatorward of the toroidal field node and closely adjacent singular points, but is coupled to an energetically neutral pattern of both velocities and magnetic fields on the poleward side. As in Gilman and Fox, we find narrow-latitude bands of sharp changes in both differential rotation and toroidal magnetic field that migrate toward the equator with increasing field strength, but are bounded in their migration by the latitude of the toroidal field node. These sharp changes are always at the locations of the singular points of the system and represent narrow domains where both kinetic and magnetic energy are being extracted from the reference state to drive the instability. We interpret the instability as a form of resonant overreflection between singular points, analogous to what happens in stratified shear flow, as described for example by Lindzen. The instability may contribute to determining the latitudinal and longitudinal distribution of active regions and other large-scale, magnetic features on the Sun, as well as enable a degree of synchronization of the evolution of the solar cycle between low latitudes and high, and between north and south hemispheres.

Influence of the hall effect on plasma compressibility in a strong toroidal magnetic field

The development of the tearing instability is studied in the presence of a high toroidal magnetic field and a high plasma conductivity. The variation of the plasma density is shown to be significant in this case.

Characterization of plasma parameters in shaped PBX-M discharges

The Princeton Beta Experiment-Modification (PBX-M) was run both with elliptical and with bean-shaped plasmas during the 1992 and 1993 operating periods. Two deuterium-fed neutral beams were used for auxiliary heating, and during 1992 the average power was . This will be referred to as the lower neutral-beam power (LNBP) period. As many as four deuterium-fed neutral beams were used during 1993, and the average power was . This will be referred to as the medium neutral-beam power (MNBP) period. The neutron source strength, Sn, showed a scaling with injected power , for both the LMBP and MNBP periods. A much wider range of shaping parameters was studied during the MNBP as compared with the LNBP period. A weak positive dependence on bean shaping was observed for the LNBP, and a stronger positive dependence on shaping was observed for MNBP, viz . High values of Sn were obtained in bean-shaped plasmas for the highest values of at for the LNBP. For the MNBP the highest values of Sn and stored energy were obtained at , and the highest values of were obtained at . The achievement of high Sn is aided by high neutral-beam power, high toroidal field, strong shaping, high electron temperature, and broad profiles. The achievement of high is aided by low toroidal field, high density, less shaping, broad profiles, and access to the H-mode, viz . The achievement of high is aided by strong shaping, high density, broad profiles, and access to the H-mode, viz . Some comparisons with the previous higher neutral-beam (HNBP) period in 1989 are also made.