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Plasma turbulence plays a key, governing role in determining the spatial-temporal evolution of plasmas in astrophysical, geophysical and laboratory contexts. In particular, turbulence on disparate spatial and temporal scales limits the level of confinement achievable in magnetic confinement fusion experiments and therefore limits the viability of sustainable fusion power 1 . The TDoTP project* (Turbulent Dynamics of Tokamak Plasmas) aims to advance understanding of plasma turbulence coupling across a wide range of scale sizes. As part of this effort, a mm-wave based scattering diagnostic is being designed for the MAST-U spherical tokamak, to measure binormal oriented high-k (electron scale) turbulence in the core plasma. We present the results of Gaussian wave optics and beam-tracing calculations 2 that demonstrate the predicted spatial and wavenumber resolution of the diagnostic, as well as estimates for the sensitivity of measurement. Current specifications for the diagnostic include an operating frequency of 260GHz, a turbulence wavenumber measurement range of k ⊥ ρ e = 0.1 -> 0.5 (where k ⊥ is the poloidal turbulence wavenumber and ρ e the electron gyroradius) and a minimum spatial localisation of ~3cm along the primary beam path.

A multi-frequency electron cyclotron resonance heating (ECRH) and current drive (CD) system is currently under construction at the ST40 spherical tokamak. The system employs 2 GYCOM gyrotrons with a maximum output power of 1 MW each and a pulse length of 2s. Both gyrotrons can be tuned to operate either at 105 GHz or 140 GHz. The system is designed to study a non-inductive plasma start-up, current ramp-up and sustainment.

Recently, 28GHz, 400kW gyrotrons have been built and tested at IAE-CAEP. The gyrotrons were developed for ECRH experiments with a spherical tokamak. The gyrotrons have a single anode magnetron injection gun to produce gyrating electron beam. The operational cavity mode is TE8,3, which is converted into a Gaussian mode by an internal quasi-optical mode converter. In order to enhance the overall efficiency, single stage depressed collectors are used. In the test, output frequency of the gyrotron is 28.1 GHz. With the cathode voltage -52kV, anode voltage about 20 kV, excited by different beam currents, the gyrotron (#2) has realized long pulse, high power output of 200 kW / 5 s, 300 kW / 3 s, and 400 kW / 1 s. Another tube (gyrotron #3) has been tested in May and delivered 400 kW / 5 s output power on May 21, 2021. Now the gyrotrons have been installed on the spherical tokamak and successfully used as trial products in the ECRH experiment. Further test of several optimized gyrotrons will continue during the next several months.

Two-cascade plasma shape control system with decoupling was designed for a linear model of a vertically elongated tokamak Globus-M2. The inner multivariable cascade controls currents in poloidal field coils, outer multivariable cascade controls plasma shape. In both cascades a decoupling matrix is used. The idea of decoupling is in reversing of matrix relationship between inputs and outputs of the system in steady-state regime. PID-controllers in the system were tuned by the QFT approach (Quantitative Feedback Theory). The obtained control system with the linear plasma model and non-linear models of current invertors as actuators for plasma position control is simulated in MatLab/Simulink environment.

A collective scattering method using the sub-THz gyrotron radiation is planned to apply to the direct detection of the electron Bernstein wave in the QUEST where the electron Bernstein wave plays one of the main heating/current drive roles in sustaining steady state spherical tokamak configuration.

In modern vertically elongated tokamaks it is critically important to maintain plasma near the first wall at the diverter phase of a plasma scenario. The reason for this is to allow plasma equilibrium reconstruction in real time, based on the external magnetic measurements of the plasma. This paper suggests a simulation approach to obtain plasma position, current, and shape control, with a plasma equilibrium reconstruction code used in feedback. The new basic idea of such a simulation is to combine scenario signals with the data coming from the linear model based on tokamak experimental data. This sum comes to the reconstruction code, which reconstructs the separatrix location and gaps between the tokamak first wall and the separatrix itself. Using the example of the functioning spherical tokamak Globus-M (Ioffe Institute, S-Petersburg, Russia), we show the results of simulation of plasma gaps in a closed-loop control system with an original H ∞ LPV plasma shape controller at upper X-point of the plasma magnetic configuration. The simulation approach may be applied to any tokamak elongated in the vertical direction to adjust plasma shape control systems before application in experiments.

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 low aspect ratio spherical tokamak (ST) MEDUSACR (Madison EDUcation Small Aspect ratio tokamak) is currently being re-commissioned in Costa Rica. The spherical tokamak was donated to the Instituto Tecnologico de Costa Rica by University of Wisconsin-Madison, USA. In addition to training, the major objective of the device ST MEDUSA-CR is to address relevant physics for spherical and conventional tokamaks. This work provides information about the re-design of the vacuum system, the new gas injection system, and the power circuit current controller of the coils, with the intention of bringing it back into operation for the investigation of plasma physics.

Active control of the toroidal current density profile is among those plasma control milestones that the National Spherical Tokamak eXperiment - Upgrade (NSTX-U) program must achieve to realize its next-step operational goals, which are characterized by high-performance, long-pulse, MHD-stable plasma operation with neutral beam heating. In this work, a previously developed physics-based control-oriented model is embedded in a feedback control scheme based on a model predictive control (MPC) strategy to track a desired current density profile evolution specified indirectly by a desired rotational transform profile. An integrator is embedded into the standard MPC formulation to account for various modeling uncertainties and external disturbances. The neutral beam powers, electron density, and total plasma current are used as actuators. The effectiveness of the proposed MPC strategy in regulating the current density profile in NSTX-U is demonstrated in closed-loop nonlinear simulations.

Regulation in time of the toroidal current profile is one of the main challenges toward the realization of the next-step operational goals for the National Spherical Tokamak eXperiment — Upgrade (NSTX-U). In this work, a nonlinear, control-oriented, physics-based model describing the temporal evolution of the current profile is first obtained by combining the magnetic diffusion equation with empirical correlations obtained for the electron density, electron temperature, and non-inductive current drives in NSTX-U. The proposed model is then embedded into the control design process to synthesize a time-variant, linear-quadratic-integral, optimal controller capable of regulating the safety factor profile around a desired target profile while rejecting disturbances. Neutral beam injectors, electron density, and the total plasma current are used as actuators to shape the current profile. The effectiveness of the proposed controller in regulating the safety factor profile in NSTX-U is demonstrated in closed-loop nonlinear simulations.

It has recently been shown that energy gain in a tokamak fusion device is, surprisingly, almost independent on device size, and is strongly dependent on energy confinement time. Coupled with recent technical advances - high temperature superconductors offering compact high performance magnets, and the Spherical Tokamak which offers high efficiency and possibly improved energy confinement- it could be possible to achieve high fusion gain in a small device. This paper describes how an independent company, Tokamak Energy, plans to exploit this new understanding, and speed the realisation of fusion energy by a series of experimental demonstrations which should secure further investment.

In ITER and future fusion reactors, intrinsic rotation will play a key role in stabilizing large scale instabilities, formation of transport barriers and reduction of divertor heat flux in the absence of significant NBI induced rotation at high injection energy. In order to utilize such intrinsic rotation property of plasma for the beneficial of a fusion grade reactors, one must understand the origin and external controlling parameters of this rotation dynamics. ECRH being one of the most indispensible and flexible heating device in tokamak, remains an attractive choice to initiate such rotations. In this paper we describe the spontaneous toroidal rotation of plasma in spherical tokamak QUEST with the help of ECRH. Several vertical magnetic field (B z ) configurations with varying mirror ratio (M) [1] (a measure of field curvature) is applied and evolution of rotation is studied with the help of Doppler spectroscopy of bulk and impurity ions. Significant toroidal rotation (V φ ∼ 20 km/s) is initiated in the open magnetic field configuration even in the initial plasma breakdown phase, which is later sustained in closed magnetic field configuration in the steady state. Rotation velocity is primarily along co-current direction and is proportional to the B z strength and resulting plasma current. High M and B z are demonstrated to be two specific external controls by which, rotation can be initiated in plasma. Response to external gas puff perturbation is investigated as a function of density and transient rotation reversal from co to counter current direction is observed. Rotation dynamics is studied in several different equilibrium configurations like limiter, single null and natural divertor IPN equilibrium [2] in QUEST and the details will be presented in this paper

Electron Bernstein wave (EBW) can be effective for heating and driving currents in spherical tokamak plasmas. Power can be coupled to EBW via mode conversion of the extraordinary (X) mode wave. The most common and successful approach to study the conditions for optimized mode conversion to EBW was evaluated analytically and numerically using a cold plasma model and an approximate kinetic model1,2. The major drawback in using RF waves was the lack of continuous wave sources at very high frequencies (above the electron plasma frequency), which has been addressed. A future milestone is to approach high power regime, where the nonlinear effects become significant, exceeding the limits of validity for present linear theory. Therefore, one appropriate tool would be particle in cell (PIC) simulation. The PIC method retains most of the nonlinear physics without approximations3. In this work, we study the direct X-B mode conversion process stages using PIC method for incident wave frequency f 0 = 15 GHz, and maximum amplitude E 0 = 105 V/m in the national spherical torus experiment (NSTX). The modelling shows a considerable reduction in X-B mode conversion efficiency, C modelling = 0.43, due to the presence of nonlinearities. Comparison of system properties to the linear state reveals predominant nonlinear effects; EBW wavelength and group velocity in comparison with linear regime exhibit an increment around ∼ 36% and 17%, respectively4.

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration (D.D. Ryutov, Phys. Plasmas, 14, 064502, 2007) may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 – 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. The radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).

By reducing neutral recycling and providing a low-Z plasma-facing surface, lithium wall coatings have led to major performance improvements in TFTR1, CDX-U2, and NSTX3. The effects of lithium coatings on plasma performance are being studied in the Lithium Tokamak Experiment (LTX), a modest size (R=0.4 m, a=0.26 m) spherical tokamak. In LTX, lithium coatings can be applied to close-fitting stainless steel surfaces surrounding ∼90% of the plasma. The surfaces can be heated to ∼300°C, allowing plasma operations with partly or fully liquid lithium surfaces.

The design features developed for the Spherical Tokamak (ST) in the PPPL pilot plant study was used as the starting point in developing designs to meet the mission of a Fusion Nuclear Science Facility (FNSF) considering a range of machine sizes based on the influence of tritium consumption and maintenance strategies. The compact nature of a steady state operated ST device for this mission pushes operating conditions and places challenges in the design of components, device maintenance and the integration of supports and services. This paper reviews the general arrangement, design details and maintenance strategy of the ST-FNSF device core for a 1.6-m and 1.0-m device; operating points which bracket the region between purchasing and breeding tritium.

The maximum achievable tritium breeding ratio (TBR), the dose to the insulator of the Cu coils, and the radial build definition are among numerous design issues investigated in detail for a Fusion Nuclear Science Facility (FNSF) based on the spherical tokamak (ST) concept. The ongoing PPPL study is considering a range of machine sizes with 1-2.2 m major radius. Preliminary shielding analysis for the PF coils of the intermediate size machine (R~1.7 m) indicated excessive dose to the cyanate ester/epoxy organic insulator, suggesting a more radiation resistant ceramic insulator such as MgO. The 3-D analysis predicts a TBR of ~1 when the details of the dual-cooled LiPb blanket and outboard penetrations are included in the model. Potential means to increase the TBR were investigated. Understanding the impact of various device sizes on the TBR is another important ongoing research activity to determine the threshold in device size for achieving T self-sufficiency.

Equilibrium simulations of spherical tokamak bean-shaped plasmas in natural divertor geometry with an Ohmic regime are presented for the first time. Preliminary results in a non-self-consistent scenario suggest higher (factor up to 2) beta values can be attained for these plasmas compared to the more usual spherical tokamak natural divertor D-shaped geometry. This can be attained simultaneously (but independently) with setting an edge localised ergodization as observed via a Poincare mapping, using a low level of perturbation to the plasma current (~2 percent). In combination with the closer proximity of the bean-shaped plasmas to the in-vessel limiters (also observed here) this leads to a higher beta limit which extrapolates favorably for a more compact reactor based in the spherical tokamak concept which is thus more economical and has a more even thermal power load.

The Mega Amp Spherical Tokamak (MAST) is the centre piece of the UK's fusion research programme. In 2010, a MAST upgrade programme was initiated with 3 primary objectives, namely to contribute to: 1) Testing of reactor concepts (in particular exhaust solutions via a flexible divertor allowing Super-X and other extended leg configurations), 2) Adding to the knowledge base for ITER (by addressing important plasma physics questions and developing predictive models to help optimise ITER's performance) and 3) Exploring the feasibility of using a spherical tokamak as a fusion Component Test Facility (looking at start-up, current drive, steady state behaviour, handling of high heat flux, plasma confinement, high beta operation and performance reliability). With the first phase of the MAST upgrade programme well under way, progress and engineering challenges will be reported in the areas of design, R&D and procurement. In particular, this will include the engineering challenges and solutions for the new divertor, the development of upgraded Toroidal Field (TF) coils to increase the toroidal field by a factor of 1.5, the approach adopted for the vertical position control (critical for this predominantly double-null device), the proposed auxiliary heating systems and the extensive diagnostic suite being developed to maximize the scientific output from the upgraded machine.

In their pioneering paper [1] Stambaugh et al developed the Peng-Hicks concept [2] of a fusion reactor based on a solid copper centre-post Spherical Tokamak (ST). Using the promising results from the START experiment [3], they produced a vision for a path to Fusion Power. This path had two elements: the ability to produce high fusion gain from an ST, and equally importantly, the ability to demonstrate this in a small (and therefore low cost) pilot plant device. In this paper we review various attempts to pursue this vision, and try to elucidate why success has not yet been achieved. However we show that the advent of high temperature superconductors (HTS) may overcome some of the problems, and we suggest a revised version of the small, low entry cost route to Fusion Power.

A compact (R 0 ∼1.2–1.3m), low aspect ratio, low-Q (

For 60 years fusion research has been focused on Fusion for Energy (F4E) as the ultimate carbon-free solution to the world's energy problems. It is proving a worthy but difficult task. However it is relatively easy to produce high-energy fusion neutrons. The many potential applications of a 14MeV neutron source are outlined, and a range of existing designs for such a source, based on a D-T fuelled Spherical Tokamak (ST), are reviewed. It is shown that the problems of high build and operating costs and uncertainties in operating conditions, can be eased by a small device SCFNS (Super-Compact Fusion Neutron Source) of major radius ∼0.5m, which although operating at modest plasma performance can provide megawatt-level neutron output. This break-through is achieved via the effectiveness of beam-plasma fusion, which becomes dominant in these conditions. Such a device would provide a resolution of the uncertainties in fusion STs (such as start-up, ramp-up, and steady-state operation); be an effective neutron source for research, and be an ideal entry vehicle for development of more powerful neutron sources in the new objective of ‘Fusion for Neutrons’ (F4N).

A fusion pilot plant study was initiated to evaluate the potential benefits of following the fission development path as an approach for the commercialization of fusion. In such an approach, a fusion pilot plant would bridge the development needs in moving from ITER to a first of a kind fusion power plant. The pilot plant mission would encompass the component test and fusion nuclear science missions yet produce net electricity. In the first phase of the study scoping designs were developed for three different magnetic configuration options: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). Critical component features have been added to the designs that impact the general arrangement and maintenance characteristics of each device. The requirements specified in defining the pilot plant challenge the machine configurations developed for each option. Developing multiple options with a consistent set of requirements enables a uniform comparison of configuration and component issues that drive each design. This paper will provide an engineering design overview of each option, address open issues and assess where further work is needed to meet the pilot plant objectives.

spherical tokamak currently operating to investigate the low recycling regime for magnetically confined plasmas through the utilization of liquid lithium coated plasma facing surfaces. The LTX machine is unique in that it incorporates inside the vacuum vessel a heated conducting shell that encloses about 80% of the last closed flux surface of the plasma. The conforming shell operates at temperatures up to 500°C and is coated with a thin liquid lithium film. Two retractable crucible evaporators are used to deposit the lithium film on the shell inner surfaces. Various methods and procedures for lithium deposition onto the shell surfaces have been tested by varying the component temperatures and vessel pressure, and utilizing concurrent glow discharge. Operating procedures and safety systems have been developed and implemented to ensure the safe operation with lithium at elevated temperatures. The machine has successfully operated with lithium coated shells demonstrating improved plasma performance as a result. Maintaining an active lithium surface between lithium evaporations was an issue and new features are currently being installed to address this. The machine has also been vented and the internal surfaces cleaned without any difficulty. Operating results, current status, ongoing upgrades and future plans will be presented.

MAST, alongside other spherical Tokamaks (STs), provides new perspectives on Tokamak physics for ITER and beyond and a platform to explore the potential of the ST as the core of a Component Test Facility (CTF) to test and develop components and technology for fusion power plants. MAST is one of the two largest STs in the world, the other being NSTX (PPPL, U.S.). Recent physics results include new measurements of the edge pedestal, the use of magnetic perturbations to influence edge-localized modes, pellet fueling, H-mode access, core transport, confinement scaling, fast-particle physics, and MHD instabilities. The program makes use of a set of unusually high resolution and wide-viewing diagnostics on MAST, since, in most experiments, the focus is on understanding the mechanisms, to support predictive modeling and comparison with other types of Tokamak. The first stage of a major upgrade is under way which will raise the heating power and toroidal field by 50% and almost double the inductive flux swing to allow long pulses and sustained noninductively driven plasmas. In particular, 17 poloidal field coils and an in-vessel cryopump will be added to allow an expanded divertor based on the “super-X” concept, as well as a conventional configuration. These upgrades will transform the operating space, pulselength, and flexibility of MAST for goal-oriented research and as a user facility. The heating and pulselength upgrades are aimed at testing off-axis current drive and fast-particle effects (important for sustained Tokamak fusion plasmas) and developing steady-state scenarios for a CTF. The divertor development is part of a growing international effort to find solutions for the power exhaust which ease the technology and material advances needed for the divertor plasma-facing components in a fusion power plant.

Centimetre wavelength broadband amplifiers at MW power levels are required for a number of applications including RaDAR as well as spherical tokamak fusion plasma heating and diagnostics. Recent publications1,2,3 on gyro-amplifiers have shown exceptional gain, power, bandwidth and efficiency performance in this wavelength range, however, when larger intrinsic and tunable bandwidths are required, gyro-amplifiers are not the best choice. We present results from a reversed guide magnetic field Free Electron Maser (FEM) amplifier experiment at Strathclyde University4. FEM oscillators5,6 have also been successfully operated at Strathclyde University but the focus of this present report is the FEM amplifier experiment. The FEM amplifier has been designed for zero beam-wave slippage to achieve maximum instantaneous bandwidth. The tunability is further extended by adjustment of beam voltage and undulator field strength. A two stage depressed collector was used to improve the amplifier efficiency.

The spherical tokamak (ST) is a leading candidate for a fusion nuclear science facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the U.S. actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5s to 5s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high β N , and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up (from I P = 0.4MA up to 0.8–1MA) as needed for an ST-based FNSF. In boundary physics, NSTX and higher-A tokamaks measure an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavorably impact next-step devices. Recently, NSTX has successfully demonstrated very high flux expansion and substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade are described.

In a number of applications, wideband amplifiers1–5 provide significant advantages over oscillators of similar output power. For instance, in spherical tokamak fusion plasma heating, frequency agility allows targeted heating of specific plasma layers, possibly providing an additional plasma stabilization technique. Experimental and theoretical investigations of wideband amplifiers at MW power levels in the centimeter to millimeter wavelength range have shown that devices with greater than 20% bandwidth, suitable for long pulse to CW operation are a practical proposition. In cases where the ultimate in amplifier bandwidth is required, reversed guide magnetic field Free Electron Maser (FEM) amplifiers5 have demonstrated tunability over 70% of center frequency with 40% instantaneous bandwidth. Gyro-amplifiers1–4 using helically corrugated waveguides6 have shown exceptional gain, power, bandwidth and efficiency performance at cm and mm wavelengths. In all cases, the performance of a long pulse (and therefore high vacuum) system is strongly influenced by factors other than the intrinsic bandwidth of the interaction. For example, the stability of all of these devices can be greatly enhanced by appropriate control of the radiation polarization. However, polarization converters work optimally at only one frequency with declining efficiency at frequencies far from the center frequency. Also low reflection windows with 40% bandwidth for circularly polarized radiation require innovative design solutions. When the design must also be compatible with 500°C bakeout it becomes even more difficult.

In a spherical tokamak (ST), an operating frequency of the reflectometry is low due to the low magnetic field of the device. In the large ST, the reflected wave signal level is weak in the large propagating length, and a direct coupling between the launching and receiving antennae may become comparable level to the reflected signal due to the diffraction effect in the low operating frequency at the ST. New single-side-band heterodyne differential-phase reflectometry is proposed on a large ST device of the QUEST, to remove the direct coupling effect in an amplitude modulation reflectometry.

The original beam injection system used in the Mega Amp Spherical Tokamak (MAST) at Culham was unable to deliver the long pulse plasma heating required to fully explore the potential of the spherical tokamak. A decision was taken to upgrade the neutral beam heating system and install two Positive Ion Neutral Injectors (PINIs), delivering 2.5 MW of deuterium neutral beam power each, when operated at 75kV and 65A.

Experimental results have shown that the high harmonic fast wave (HHFW) at 30 MHz can provide substantial plasma heating and current drive for the NSTX spherical tokamak operation. However, the present antenna strap design rarely achieves the design goal of delivering the full transmitter capability of 6 MW to the plasma. In order to deliver more power to the plasma, a new antenna strap design and the associated coaxial line feeds are being constructed. This new antenna strap design features two feed-throughs to replace the old single feed-through design. In the design process, CST Microwave Studio has been used to simulate the entire new antenna strap structure including the enclosure and the Faraday shield. In this paper, the antenna strap model and the simulation results will be discussed in detail. The test results from the new antenna straps with their associated resonant loops will be presented as well.

TST-2 is a spherical tokamak (ST) device, and High Harmonic Fast Wave (HHFW) injection experiments were performed to heat electrons in the spherical tokamak. In TST-2, electron heating was observed in some discharges, but when Parametric Decay Instability (PDI) occurred the injected wave energy is transferred to two daughter modes. PDI during HHFW injection in the NSTX ST device creates modes, which are presumably an Ion Bernstein wave (IBW) and a quasi mode at ion cyclotron frequency (ICQM). A similar decay was observed for low toroidal magnetic field discharges in TST-2. For higher toroidal magnetic field discharges, we found another type of decay, in which another peak in the frequency domain was detected between a pump wave (i.e., HHFW) frequency and the IBW frequency. Experiments are performed to identify this new peak. RF pickup probes are used to measure the magnetic field oscillations outside the closed magnetic surface region. Major results are as follows. The frequency of the decay waves coupled with ICQM increases with the toroidal magnetic field as expected. When Bt varies from 0.13 to 0.18 T, the frequency difference of the decay wave and the pump wave increases from 1.5 to 1.9 MHz in proportion to Bt- The new unidentified wave shows a similar dependence, and its frequency difference varies from 0.54 to 0.70 MHz in proportion to Bt- The width of this new peak is broader than that of IBWs peak. These results suggest that the new peak corresponds to the decay wave coupled with Alfven mode, because the fact that the frequency of Alfven mode is in proportional to the magnetic field, is consistent with the experimental results.

In Saint-Petersburg state University the modernization of spherical Gutta tokamak is carried out. The aim of the program is to build the research device to study plasma position, shape and current control problems in the tokamak, as well as to assist the scientific and educational programs on various plasma physics problems. Such feature of this device as a relatively small thickness of the vacuum vessel wall made as momentless shell allows changes in the magnetic field in plasma region, to measure the plasma parameters with a minimum delay using electromagnetic diagnostics and changes in toroidal field as necessary.

The MAST (mega amp spherical tokamak) fusion research facility at Culham is undergoing an upgrade of its neutral beam heating systems. The two neutral beam injection systems (NBI), presently installed, can each deliver 2 MW of power for a period of /spl sim/300 ms. These injectors will be upgraded to a longer pulse capability by the installation of new ion sources. These will be a variation of the Joint European Torus (JET) 80 kV tetrode positive ion neutral injector (PINI) design with optics customized to suit the MAST duct geometry. They will eventually be capable of pulse lengths of up to 20 seconds with an injectable total power of 2.5 MW / beamline. In anticipation of these enhancements, several beamline components have been upgraded to cope with the increased beam pulse lengths delivered by the new PINI sources. In particular, the residual ion dumps (RIDs) and calorimeters have been redesigned to use actively cooled hypervapotrons as beam stopping elements. These design solutions are discussed in the paper. Hypervapotrons have been used reliably at JET for many years, their thermal performance being optimized to suit specific needs. Their application to this function has required a further enhancement in performance. The changes made and the technical justification for them is discussed along with the results from the subsequent full-scale power handling tests. A new high power handling record for hypervapotrons of 20 MW/m/sup 2/ was established in these tests. A large-scale mock-up of the internal geometry of the hypervapotron was constructed so that flow visualization could be used to confirm the assumed flow patterns within the hypervapotron. High-speed digital video was used to enable the examination of the flow pattern within the cooling channels. The high-resolution film of the seeded flow was then analyzed at low speed with a view to gaining a better understanding of the heat transfer mechanism. The results of this examination are presented.

Burning criteria for a D-3He fusion reactor have been investigated by using the 0-D power and particle balance equations and the IPB98(y,2) scaling law. A spherical tokamak (ST) configuration with the internal transport barrier (ITB) and the bootstrap current fraction of 100 % are assumed. As main results, the hot ion mode (Ti/T e>1) and the high central beta are essential to realize a D-3He fusion reactor, and the confinement improvement does not reduce the requirement of the hot ion mode. The reactor size, the magnetic field, confinement improvement factor and the heat flux may be attainable if the hot ion mode is realized

Summary form only given. Recent progress on spherical tokamak (ST) experiments has verified its advantage of high beta capability, which provides attractive engineering benefits connected with a compact design to realize an economical fusion power plant. Because of a severely restricted space at the narrow center column of ST devices, removing the central ohmic solenoid is considered to be a crucial step for a future plant and is also a key requirement of CTF. It is therefore strongly desired to develop an alternative way to ohmic heating to initiate and ramp-up the plasma current to a level high enough for the second heating and current drive by NBI to reach ignition. The main objective of LATE (low aspect ratio torus experiment) is to demonstrate the formation of ST by electron cyclotron heating (ECH) alone without the center solenoid. By injecting a 2.45 GHz microwave pulse up to 30 kW for two seconds, a plasma current has been initiated and ramped-up slowly to 7.2 kA with a slow ramp-up of the vertical field to maintain the equilibrium of the plasma current loop. Firstly, a steady toroidal field and a steady vertical field (Bv = 15 Gauss at R = 27 cm) are applied and hydrogen gas is introduced. Microwaves from three magnetrons are injected from outboard side with the linearly polarized electric field on the equatorial plane. An initial plasma is instantly produced at the fundamental EC resonance layer, and quickly expands into the low field side. A plasma current is then generated and increased up to 2.0 kA spontaneously and an initial closed flux surface is formed under the steady vertical field. The current subsequently ramps up slowly by increasing the microwave power and the Bv field in order to keep the higher current in equilibrium, and finally reaches Ip = 7.2 kA at Bv = 78 Gauss. This value of the Ip amounts 12 percent of the toroidal field coil current. Magnetic analysis shows that an ST equilibrium, having the last closed flux surface with an aspect ratio of R0/a = 20.5 cm/15 cm ~ 1.4, an elongation of kappa = 1.6 and qedge = 34, has been formed and maintained at the final discharge stage. The plasma current center is kept near the 2nd EC resonance layer and the line averaged electron density significantly exceeds the plasma cutoff density, suggesting that the second harmonic ECH by the mode converted electron Bernstein wave could be responsible for heating in the present plasma

Summary form only given. The characteristic low magnetic field and high plasma density of a spherical Tokamak does not permit the typical radiation of O and X modes from the first five electron cyclotron harmonics. Thus only electron Bernstein waves, (modes not subject to a density limit) can be responsible for ECE radiation upon which mode convert into electromagnetic waves in the upper hybrid resonance region can be responsible for ECE radiation. A detailed 3 D model of the MAST equilibrium plasma is determined from the EFIT code and high spatial resolution Thomson scattering data. Taking into account the horn, thin lens two mirrors and a window associated with the RF antenna, a comparison is made between measured ECE and theoretical predictions from a Gaussian 3 D ray tracing code. For each ray we determine the spot on the separatrix and also the corresponding EBW-X-O mode conversion efficiency in the transport barrier. Using the reciprocity between EBW emission and EBW absorption, the EBW rays are launched from the upper hybrid resonance (UHR) towards the plasma center. Since N/spl par/ strongly oscillates for rays launched near the equatorial plane, nonlocal absorption characteristics must be considered. The effective radiation temperature is determined by solving the ray equations coupled to the radiative transfer equation, using the Rayleigh-Jeans black body radiation law. At oblique incidence, the antenna beam is partially blocked (and reflected) by the finite size of the window. This, and similar geometric effects, are responsible for the enhancement of the low frequency part of the ECE spectra (larger visible area) and suppression of the higher frequency parts of the spectrum (blocked by the window). Taking these effects into account, the comparison between theory and experimental data is good.

Summary form only given, as follows. We show that the SOL statistical properties are similar in four extremely different confinement devices, that are, PISCES linear plasma device, the Tore Supra tokamak, the Alcator C-Mod tokamak and the MAST spherical tokamak. This is performed for the probability distribution function, the power spectra, and using the conditional averaging technique.

The Mega Amp Spherical Tokamak (MAST) is an experimental fusion device currently operating at the Culham Science Centre. The low aspect ratio of the spherical tokamak combined with the need to develop a high flux swing from its central solenoid leads to high forces on the conductors that form the centre column. This column consists of a compact solenoid wound around the inner limbs of the toroidal field coils. In order to achieve the nominal plasma parameters, the dimensions of the centre column were optimised to make best possible use of the space available. The most critical region of the solenoid magnet design is the end turn and tail section where complex 3D magnetic fields interact with the current in the end turn conductors giving an asymmetric stress distribution. The hoop load in this end turn is particularly troublesome since it must be reacted across a bonded joint, which inevitably generates tensile and shear stresses within it. This paper describes how this problem has been solved for the new MAST centre column by the use of compact quadrupole tails formed close to the main coil and connected to it by tight bends formed in the water cooled copper conductor. A test coil employing these features and fitted with engineering diagnostics has been manufactured and tested at Culham to the maximum design current. This coil not only demonstrated the manufacturing processes involved but also provided test data, which compares well with engineering analysis.

The MAST (Mega-Amp Spherical Tokamak) experiment has been operational since Jan 2000. Results from MAST are important both in evaluating the potential of future ST fusion devices, and in developing understanding of processes relevant to conventional aspect ratio tokamaks. In this paper methods of initiating, ramping up, and sustaining the high plasma currents associated with ST fusion devices are discussed. New physics results, including the effectiveness of inboard gas puffing, the threshold power required for the L-H transition, and the 'natural' divertor configuration are given. Plans for improvements during 2002 are outlined.

Plasma physics and technology features of the low-aspect-ratio spherical tokamak (ST) combine to determine the characteristic performance of a projected DT-fueled, 1000-MWe-class, commercial fusion power plant, designated ARIES-ST. Key attributes of the final design point are summarized.

The NSTX device plasma facing components [PFC] consist of inboard divertors, outboard divertors, primary passive plates, secondary passive plates, a center stack casing (CSC), and the heating/cooling fluid distribution system. The PFC surfaces are protected by 3584 individually mounted carbon tiles. Surfaces exposed to high heat flux and/or high loads utilize composite C-C graphite and the remainder utilize less costly ATJ graphite. A variety of diagnostics are incorporated into the PFCs including thermocouples, Langmuir probes, Mirnov coils, and Rogowski coils. The NSTX device is designed to be operated in a pulse mode of five seconds on followed by five minutes off. Its PFC components are also required to be baked out to 350 /spl deg/C. During operation the PFC tiles are permitted to ramp up thermally and then cool sufficiently between shots to prevent ratcheting during subsequent shots. The CSC tiles are required to be thermally isolated from the CSC so that the primary heat loss is radiation to the other PFC components. The other PFCs are thermally coupled to water cooled plates by conductive gaskets. Special mounts are required which permit thermal expansion and can withstand disruption loads while maintaining thermal contact. The tiles and mounts for the CSC are required to fall within a total radial space allotment of only 14 mm. A unique design for mounting graphite tiles to the CSC was developed which utilizes drift (shear) pins and Inconel brackets. Installation is accomplished via hidden fasteners accessed through very small holes in the tile faces. Analyses of the CSC mounting structure were performed and pull tests were performed on assemblies which simulated the attachment geometry in an attempt to determine the ultimate strength of the configuration and the mechanism of failure.

ARIES-ST, a fusion power plant design based on the spherical tokamak concept, has many attractive features, including high beta and power density, low magnetic field and high self-driven current. The physics basis for choosing the design plasma parameters are described with respect to MHD equilibrium and stability, current drive and startup. Implications for the fusion power core design are highlighted, and key directions for ST physics research are identified.

Summary form only given. In this report, we present the results on the novel ultra low aspect ratio stellarator system with A=1.05-2, that can be called a spherical stellarator (SS) in analogy with the spherical tokamak (ST) concept. Coil configuration of SS features the central stack and differs from that of ST by inclination of external parts of the toroidal field coils. A few potential advantages of the SS configuration, over other approaches in magnetic confinement fusion, can be summarized as follows: (a) very compact design offering the low cost approach, (b) closed vacuum flux surfaces and no need, in principle, in the ohmic current transformer, (c) simple modular coil system, (d) easy access to the plasma, (e) more area for heat removal and the tritium breeding blanket in a reactor, (f) simple toroidally symmetric divertor, (g) enough space between the central stack and the plasma surface to put the blanket and protect the stack from intense fluxes of particles, heat and neutrons.

Summary form only given. The upgraded Helicity Injection Tokamak (HIT-II) extends coaxial helicity injection (CHI) to time scales longer than the wall resistive diffusion time. Analysis of HIT data and stability studies have developed a proposed mechanism for CHI current drive. This mechanism, consistent with these analyses, appears to be responsible for current drive inside the separatrix, and is summarized as follows: a hollow current profile plus a small (2-10 mm) edge vacuum region causes an n=1 external kink mode to go unstable at the edge; on a resistive diffusion time scale the equilibrium adjusts to this perturbation throughout the volume, producing a global helical magnetic structure locked to the electrons and not to the ions; the externally driven electrons on the open field lines cause the structure to rotate; the rotating structure, in turn, drags the electrons in the closed flux region producing current drive inside the separatrix.

ARIES-ST, a fusion power plant design based on the spherical tokamak concept, has many attractive features, including high beta and power density, low magnetic field, high self-driven current, and a compact power core. The physics basis for three aspects of the plasma performance and operation, namely, MHD stability, current drive and profile control, and startup have been investigated during the first phase of the design, and the results to date are reported.

Globus-M spherical tokamak is to be constructed at A.F. Ioffe Institute, St. Petersburg next year. The problems of the Globus-M most critical components design and manufacturing are basically solved. At the same time the revision of the basic physical and engineering parameters in the process of tokamak design and manufacturing have resulted in a change of the device engineering concepts. The final design decisions for the magnets as well as the status of manufacture and tests of critical components are discussed in the paper.

Motivations and constraints regarding the identification of DT-fueled, 1,000-MWe-class commercial power-plant embodiments of the (low-aspect-ratio) spherical tokamak (ST) are summarized in the context of ongoing ARIES studies. Key design drivers are identified and the prospects of this approach to magnetic fusion energy are considered.