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The eight triplets of straps of the ITER ICRF antenna array are fed through 8 matching circuits and 4 hybrids to ensure load resilience. Decouplers are used to mitigate the effects of triplet mutual coupling. They also control the array phasing. The electrical constraints on the decouplers for different layouts with heating (H) or current drive (CD) phasing are compared starting from the TOPICA matrix computed for the last antenna plug design and the reference (most pessimistic) plasma profile “2010low” provided by IO. It is shown that this last profile provides a significant decrease of plasma coupling and increase of mutual coupling with respect to the previous reference profile “Sc2short17”. This results in a larger range of decoupler reactance Xdec and voltage VXdec needed. This range can be reduced when using 10 decouplers instead of the 7 needed for the same forward power PGk+ of the 4 power sources. For H phasing only 4 decouplers could be used but with different PGk+ (PGk+ ratio up to 1.5–2.5). For CD phasing and same plasma profile the power capability Ptot is increased by 25% with a decoupler layout allowing much smaller poloidal phasing than the 90° provided by the hybrids. A decrease of the distance antenna-plasma profile reduces the normalized decoupler voltage VXdec/√Ptot with no significant change of the Xdec range. The recess of the vertical septa between the strap boxes increases the plasma coupling but has the drawback of also increasing the mutual coupling between triplets: the needed range of Xdec and of VXdec/√Ptot is increased in proportion.

The ITER Ion Cyclotron Heating and Current Drive system (IC H&CD) is designed to deliver 20MW to a broad range of plasma scenarios between 40 and 55MHz, during very long pulses. It consists of two broadband equatorial port plug antennas, their pre-matching and matching systems, transmission lines, Radio Frequency (RF) Sources and High Voltage Power Supplies. The overall project schedule has been revised and agreed by ITER Council; it re-integrates the second antenna and its power supplies in construction baseline and sets the dates for progressive installation with DT phase planned in 2035. Recent progress on ICRF subsystems is reported, covering design evolution, qualification of test articles and specific R&D results in domestic agencies, suppliers, associated laboratories and IO.

The US-ITER in-kind contribution to the ICH (Ion Cyclotron Heating) system for the tokamak being built in Cadarache, France, is the coaxial transmission lines, matching systems, and matching algorithms. A prematching system composed of 20 ohm triple-stub-tuned lines was proposed to match to the antenna. Because of constraints on the port cell and space availability, triple stub prematching may not be possible, so alternative prematching systems are proposed for use. Results of initial analyses of the prematching concepts are presented.

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.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

The ITER burning plasma and advanced operating regimes require robust and reliable heating and current drive and fueling systems. The ITER design documents [1,2] describe the requirements and reference designs for the ion cyclotron and pellet fueling systems. Development and testing programs are required to optimize, validate and qualify these systems for installation on ITER. The ITER ion cyclotron system offers significant technology challenges. The antenna must operate in a nuclear environment and withstand heat loads and disruption forces beyond present-day designs. It must operate for long pulse lengths and be highly reliable, delivering power to a plasma load with properties that will change throughout the discharge. The ITER ion cyclotron system consists of one eight-strap antenna, eight rf sources (20 MW, 35-65 MHz), associated high-voltage DC power supplies, transmission lines and matching and decoupling components. The ITER fueling system consists of a gas injection system and multiple pellet injectors for edge fueling and deep core fueling. Pellet injection will be the primary ITER fuel delivery system. The fueling requirements will require significant extensions in pellet injector pulse length (∼3000 s), throughput (400 torr-L/s,) and reliability. The proposed design is based on a centrifuge accelerator fed by a continuous screw extruder. Inner wall pellet injection with the use of curved guide tubes will be utilized for deep fueling.

An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

Ion absorption of the high harmonic fast wave in a spherical torus [Y.-K. M. Peng et al., Nucl. Fusion 26, 769 (1986)] is of critical importance to assessing the viability of the wave as a means of heating and driving current. Analysis of recent National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)] shots has revealed that under some conditions when neutral beam and rf power are injected into the plasma simultaneously, a fast ion population with energy above the beam injection energy is sustained by the wave. In agreement with modeling, these experiments find the rf-induced fast ion tail strength and neutron rate at lower B-fields to be less enhanced, likely due to a larger β profile, which promotes greater off-axis absorption where the fast ion population is small. Ion loss codes find the increased loss fraction with decreased B insufficient to account for the changes in tail strength, providing further evidence that this is a rf interaction effect. Though greater ion absorption...

The two-dimensional radial vs poloidal structure and motion of edge turbulence in the National Spherical Torus Experiment (NSTX) were measured using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analysed for Ohmic, L- and H-mode plasma conditions. The two-dimensional images often show regions of strong localized light emission known as 'blobs', which move both poloidally and radially at a typical speed of ?105?cm?s?1, and sometimes show spatially periodic features.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

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 e 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.

The ion cyclotron heating and current drive system on the National Spherical Torus Experiment (NSTX) has delivered over 3 MW reliably for pulse lengths over 100 ms with various phasings of the antennas. A circuit model of the system that includes the 12 coupled antennas and six radio-frequency sources has been developed that gives good agreement with vacuum measurements. When it is used to experimentally determine the S-matrix of the system under different plasma conditions, pronounced asymmetries in the off-diagonal values of the S-matrix are seen. The S-matrix in the presence of plasma has been calculated with the RANT3D code using measured edge density profiles in front of the antenna; these agree remarkably well with the measurements. The asymmetry is caused primarily by the large pitch angle of the magnetic field in front of the antenna, coupled with the gradients in the plasma edge.

High harmonic fast wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, S. Neumeyer et al., in Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999 (IEEE, Piscataway, NJ, 1999), p. 53] is such a device. An rf heating system has been installed on the NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to ...

Edge parameters play a critical role in high confinement mode (H-mode) access, which is a key component of discharge optimization in present day toroidal confinement experiments and the design of next generation devices. Because the edge magnetic topology of a spherical torus (ST) differs from a conventional aspect ratio tokamak, H-modes in STs exhibit important differences compared with tokamaks. The dependence of the National Spherical Torus Experiment (NSTX) [C. Neumeyer et al., Fusion Eng. Des. 54, 275 (2001)] edge plasma on heating power, including the low confinement mode (L-mode) to H-mode (L-H) transition requirements and the occurrence of edge-localized modes (ELMs), and on divertor configuration is quantified. Comparisons between good L-modes and H-modes show greater differences in the ion channel than the electron channel. The threshold power for the H-mode transition in NSTX is generally above the predictions of a recent International Tokamak Experimental Reactor (ITER) [ITER Physics Basis Edi...

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

The paper reports on latest developments for the ITER Ion Cyclotron Heating and Current Drive system: imminent acceptance tests of a prototype power supply at full power; successful factory acceptance of candidate RF amplifier tubes which will be tested on dedicated facilities; further design integration and experimental validation of transmission line components under 6MW hour-long pulses. The antenna Faraday shield thermal design has been validated above requirements by cyclic high heat flux tests. R&D on ceramic brazing is under way for the RF vacuum windows. The antenna port plug RF design is stable but major evolution of the mechanical design is in preparation to achieve compliance with the load specification, warrant manufacturability and incorporate late interface change requests. The antenna power coupling capability predictions have been strengthened by showing that, if the plasma scrape-off layer turns out to be steep and the edge density low, the reference burning plasma can realistically be displaced to improve the coupling.

FIRE is a high-field, burning-plasma tokamak that is being studied as a possible option for future fusion research. Preliminary parameters for this machine are R 0 =2.14 m, a =0.595 m, B 0 =10 T, and I p =7.7 MA. Magnetic field coils are to be made of copper and pre-cooled with LN 2 before each shot. The flat-top pulse length desired is ≥20 s. Ion cyclotron (IC) and lower hybrid rf systems will be used for heating and current drive. Present specifications call for 20 MW of IC heating power, with an additional 10 MW of IC power or 20 MW of lower hybrid power as upgrade options.

A prototype antenna for the ITER ion cyclotron system, based on a resonant double loop antenna design from the 1998 Engineering Design Activity, has been built and tested. Electrical properties of the antenna were measured and were in agreement with theoretical calculations. Results of high-power experiments in vacuum showed that the antenna could operate at voltages in excess of 60 kV for multisecond pulses. The stub tuning arrangement proposed for use on ITER was demonstrated to work. These results validate the original design concept of the ITER antenna.

A high harmonic fast wave (HHFW) antenna array, designed to provide up to 6 MW of power at 30 MHz for heating and current drive applications, has been operated on the NSTX experiment at Princeton Plasma Physics Laboratory (PPPL). The full array consists of 12 evenly spaced, identical current strap modules connected in pairs. Each pair is connected as a half-wave resonant loop and is intended to be driven by one transmitter, allowing rapid phase shift between transmitters. A decoupling network compensates for the mutual inductive coupling between adjacent current straps, effectively isolating the six transmitters from one another. Initial rf operation between November 1999 and January 2000 used eight straps to form four loops, which were driven by two transmitters. Two adjacent loops were connected with a λ/2 coax section to be driven out of phase by a single transmitter. Up to 2 MW of power was delivered during this stage of operation; inter-loop phasings of 0–π–π–0 and 0–π–0–π were investigated. Models of the power distribution system indicate the nominal plasma loading was about 5 Ω/m, close to the design value of 6 Ω/m. The HHFW system has now been reconfigured for 12-strap, six-transmitter operation with decouplers; low power vacuum and plasma measurements have begun.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a approximately equal to 1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of High Harmonic Fast Wave (HHFW) heating and current drive, 5 MW of Neutral Beam Injection (NBI) and Co-Axial Helicity Injection (CHI) for non-inductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high-bt and high-bootstrap current fraction, and to develop methods for non-inductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA and with a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 msec was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and bt =21% were produced.

The Korea Superconducting Tokamak Advanced Research (KSTAR) project is the major effort of the national fusion programme of the Republic of Korea. Its aim is to develop a steady state capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. The major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T and plasma current 2 MA, with a strongly shaped plasma cross-section and double null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beams, ion cyclotron waves, lower hybrid waves and electron cyclotron waves for flexible profile control in advanced tokamak operating modes. A comprehensive set of diagnostics is planned for plasma control, performance evaluation and physics understanding. The project has completed its conceptual design and moved to the engineering design and construction phase. The target date for the first plasma is 2002.

The National Spherical Torus Experiment (NSTX) is being built at the Princeton Plasma Physics Laboratory to test the fusion physics principles for the Spherical Torus (ST) concept at the MA level. The NSTX nominal plasma parameters are R {sub 0} = 85 cm, a = 67 cm, R/a greater than or equal to 1.26, B {sub T} = 3 kG, I {sub p} = 1 MA, q {sub 95} = 14, elongation {kappa} less than or equal to 2.2, triangularity {delta} less than or equal to 0.5, and plasma pulse length of up to 5 sec. The plasma heating/current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up, as well as a dispersive scrape-off layer for heat and particle flux handling.

Ion cyclotron (IC) system for KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak has been designed to be capable of fast sweeping the phase intervals of the antenna strap currents with all the mechanical tuners fixed. It is shown that the fast sweep is possible by use of proper decoupler circuits. Design and tuning method for the decoupler circuit are described. We show that the circuit works satisfactorily by examining its behavior under the fast sweep.

Requirements and design characteristics of the heating and current drive systems for KSTAR (Korean Super conducting Tokamak Advanced Research)1,2 tokamak are described. The neutral beam (NB), the i...

Measurements of radiofrequency (RF) loading on the Doubler III D shape (DIII-D) tokamak using a fast wave ion cyclotron antenna with and without a Faraday shield present indicate that the loading resistance seen by the antenna is significantly higher at low values of RF power (P ≤ 20 kW with no Faraday shield) than is predicted by standard fast wave antenna coupling theory. Measurements of the density profile in the plasma edge region allow a direct comparison of calculated and measured loading. At high power, the calculated fast wave loading agrees with the measurements. The high loading observed at low power can be explained by direct losses in an RF enhanced sheath. The density value at the current strap of approximately=4*1010 cm-3 needed to give the observed loading behaviour is in agreement with the measured density values in the plasma scrape-off region, extrapolated to the current strap location using a simple Bohm diffusion model

Deep brain stimulation (DBS) surgery is an effective treatment for patients with advanced Parkinson's disease. Delirium in hospitalized Parkinson's disease patients is common and often leads to prolonged hospital stays. This study reports on the incidence and etiology of postoperative delirium following DBS surgery. Patients (n=59) with advanced Parkinson's disease underwent bilateral (n=56) or unilateral (n=3) DBS electrode implant surgery, followed 1 week later with surgical placement of DBS generators. The development of delirium during either hospital stay was evaluated retrospectively from the hospital chart. Potential causes of delirium were evaluated, including history of delirium, opiate equivalents, medication administration delays and missed doses during hospitalization, and Parkinson's disease duration. Delirium following implantation of DBS electrodes was common (22% of patients). It was less commonly associated with generator placement (10%). A history of delirium, age, and disease duration were positive predictors of delirium. Opiate equivalent doses were negatively correlated with delirium. Missed Parkinson's medication doses (53% of patients) and delayed administration (81% of patients) were common, and had a slight relation with delirium. Delirium was not related to complexity of medication regimen or use of dementia medications. Despite the presence of delirium most patients still only required a single night in the hospital post-surgery (67%). Prolonged hospital stay was due not only to delirium but also severe off states and other medical issues. Recognition and expectant management of delirium is best accomplished in a multidisciplinary setting, including the patient's family and nursing, pharmacy and neurological surgery staff.

The physics capabilities of an ignition-and-moderate-burn tokamak to explore the physics of burning plasmas and burn control on ash accumulation time scales are described. The machine provides physics capabilities comparable to those of the International Thermonuclear Experimental Reactor (ITER) for pulse lengths up to 120 s, but lacks the nuclear component testing, superconducting magnet technology, and long-pulse aspects of ITER`s mission. Strong plasma shaping is adopted to reduce the cost relative of ITER. Using ITER guidelines to evaluate the physics performance, this machine has the same ignition margin as ITER`s, and operates within the limits on beta, density (i.e., the Greenwald density limit), and safety-factor specified in the ITER physics guidelines. Acceptable peak heat fluxes to divertor target surfaces are maintained with an attached, high-recycling divertor operating scenario typical of present-day machines. A range of ignited and driven operating modes is available, including advanced modes prototypical of steady-state tokamak operation. 10 refs., 3 figs., 3 tabs.

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40-55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system components. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

Objective Identify and categorize recommendations made by a consultant pharmacist in a neurology clinic, and document patient satisfaction with consultations provided by the pharmacist. Design Prospective, case series. Setting Ambulatory neurology clinic located in eastern Washington state. Patients Participants were referred to the consultant pharmacist by the neurologist or neurology nurse practitioner or self-referral. The pharmacist saw 56 patients, for a total of 66 encounters. Interventions A medication review was completed, and the patient and pharmacist met for consultation, assessment, and patient counseling. The pharmacist made pharmacotherapeutic recommendations and referrals to other health care providers. Printed surveys were distributed to patients regarding their satisfaction with the pharmacy consultation. Main outcome measure(s) The number and types of recommendations made by the pharmacist. Patient satisfaction with the pharmacist and pharmacy consultation, reported in a written survey provided after the consultation. Results The pharmacist spent an average of 38 minutes with each patient. A total of 192 pharmacotherapeutic recommendations were made: 55 (29%) to discontinue a medication, 46 (24%) to add a medication, 45 (23%) to change a dose, 39 (20%) therapeutic substitutions, and 7 (4%) for therapeutic monitoring. Survey respondents indicated they were very satisfied (97%), and all (100%) had a better understanding of how to take their medications after meeting with the pharmacist. Conclusion Pharmacist involvement in a neurology clinic increases patient-pharmacist contact time and patient understanding regarding medications. The pharmacist offers numerous recommendations to health care providers, potentially improving the patient's pharmacotherapy.

The heating and current drive (H and CD) system proposed for the TPX tokamak will consist of ion cyclotron, neutral beam, and lower hybrid systems. It will have 17.5 MW of installed H and CD power initially, and can be upgraded to 45 MW. It will be used to explore advanced confinement and fully current-driven plasma regimes with pulse lengths of up to 1,000 s.

The Ion Cyclotron Heating and Current Drive (IC H & CD) system for ITER will provide Radio frequency (RF) heating and current drive to the ITER plasma. This system is designed to provide 20 MW into the plasma using frequencies in the range of 40 MHz to 55 MHz, which could be upgraded to 40 MW in future. This paper presents the recent development of the IC H&CD system configuration in the ITER Tokamak complex & auxiliary buildings with developed interfaces.

The measured penetration depth lambda of deuterium pellets injected into the Joint European Torus (JET) confirms some features of the neutral gas shielding (NGS) model, but not others. The scaling of lambda with plasma and pellet parameters agrees with the NGS model, as in earlier ASDEX studies. Pellet velocity was varied over the range 0.46-1.35 km/s in the JET experiments to test specifically the scaling of lambda with velocity. This scaling also agrees with the NGS model. However, the penetration is deeper in JET than in ASDEX when it is corrected for the expected machine size dependence. Furthermore, the penetration depths measured in JET are greater (by nearly a factor of two) than those predicted by local ablation calculations using the NGS model with an incident Maxwellian distribution of electrons. Plasma shielding used in previous modelling of the JET penetration data can account for the additional shielding, but it also removes the observed velocity dependence. The implications of both the scaling observations and the penetration depths for improvements in ablation theory and in the models are discussed

Fast wave heating experiments in the ion cyclotron range of frequencies (ICRF) were performed on target plasmas produced by 350 kW of electron cyclotron heating at 53 GHz and also by neutral beam injection in the Advanced Toroidal Facility (ATF). Various heating regimes were investigated in the frequency range between 9.2 MHz and 28.8 MHz with magnetic fields of 0.95 T and 1.9 T on axis. The nominal pulse lengths of up to 200 kW RF power were in the range between 100 and 400 ms. Data from spectroscopy, loading measurements, and edge RF and Langmuir probes were used to characterize the RF induced effects on the ATF plasma. In the hydrogen minority regime at low plasma density, large suprathermal ion tails were observed with a neutral particle analyser. At high density (ne ≥ 5.0 × 1013 cm-3) substantial increases in antenna loading were observed, but ICRF power was insufficient to produce definitive heating results. A two-dimensional RF heating code, ORION, and a Fokker-Planck code, RFTRANS, were used to simulate these experiments. A simulation of future high power, higher density experiments in ATF indicates improved bulk heating results due to the improved loading and more efficient thermalization of the minority tail

The electron cyclotron (EC), ion cyclotron (IC), heating-neutral beam (H-NB) and, although not in the day 1 baseline, lower hybrid (LH) systems intended for ITER have been reviewed in 2007/2008 in light of progress of physics and technology in the field. Although the overall specifications are unchanged, notable changes have been approved. Firstly, it has been emphasized that the HC the possibility to operate IC with 2 antennas with increased robustness (no moving parts); the possible increase to 2 MW of key components of the EC transmission systems in order to provide an easier upgrading of the EC power as may be required by the project; the addition of a building dedicated to the RF power sources and to a testing facility for acceptance of diagnostics and heating port plugs. Thirdly, the need of a plan for developing, in time for the active phase, a CD system such as LH suitable for very long pulse operation of ITER was recognised. The review describes these changes and their rationale.

FIRE is a proposed high-field tokamak (B/sub 0/ up to 10 T) designed to explore burning plasma physics of conventional ELMy H-modes as well as several types of advanced tokamak (AT) modes. A 20-MW ion cyclotron system in the 70 to 115 MHz frequency range is the baseline heating system. This system should achieve Q /spl ap/ 10 in a 50:50 D-T plasma under standard (i.e., non-AT) conditions, and can also drive a modest amount of current near the center of the plasma during AT operation. Two other heating and current drive systems are being considered for FIRE as upgrade options: a 30-MW, 5-GHz lower hybrid system to drive current in the outer part of the plasma, and a 170-GHz electron cyclotron system to stabilize neoclassical tearing modes during AT operation.

The RF source requirements for the ITER ion cyclotron (IC) heating and current drive system are very challenging- 20 MW CW power into an antenna load with a VSWR of up to 2 over the frequency range of 35-65 MHz. For the two present antenna designs under consideration, 8 sources providing 2.5 MW each are to be employed. For these sources, the outputs of two final power amplifiers (FPAs), using the high power CPI 4CM2500KG tube, are combined with a 180deg hybrid combiner to easily meet the ITER IC source requirements-2.5 MW is supplied at a VSWR of 2 at ~70% of the maximum tube power available in class B operation. The cylindrical cavity configuration for the FPAs is quite compact so that the 8 combined sources fit into the space allocated at the ITER site with room to spare. The source configuration is described in detail and its projected operating power curves are presented. Although the CPI tube has been shown to be stable under high power operating conditions on many facilities, a test of the combined FPA source arrangement is in preparation using existing high power 30 MHz amplifiers to assure that this configuration can be made robustly stable for all phases at a VSWR up to 2. The possibility of using 12 sources to feed a suitably modified antenna design is also discussed in the context of providing flexibility for specifying the final IC antenna design

Thomson scattering laser pulses are synchronized relative to modulated HHFW power to permit evaluation of the electron energy confinement time during and following HHFW pulses for both heating and current drive antenna phasing. Profile changes resulting from instabilities require that the total electron stored energy, evaluated by integrating the midplane electron pressure P(sub)e(R) over the magnetic surfaces prescribed by EFIT analysis, be used to derive the electron energy confinement time. Core confinement is reduced during a sawtooth instability but, although the electron energy is distributed outward by the sawtooth, the bulk electron energy confinement time is essentially unaffected. The radial deposition of energy into the electrons is noticeably more peaked for current drive phasing (longer wavelength excitation) relative to that for heating phasing (shorter wavelength excitation) as is expected theoretically. However, the power delivered to the core plasma is reduced consider ably for the current drive phasing, indicating that surface/peripheral damping processes play a more important role for this case.

High Harmonic Fast Wave (HHFW) heating experiments on NSTX have been observed to be accompanied by significant edge ion heating (T{sub i} >> T{sub e}). This heating is found to be anisotropic with T{sub perp} > T{sub par}. Simultaneously, coherent oscillations have been detected with an edge Langmuir probe. The oscillations are consistent with parametric decay of the incident fast wave ({omega} > 13{omega}{sub ci}) into ion Bernstein waves and an unobserved ion-cyclotron quasi-mode. The observation of anisotropic heating is consistent with Bernstein wave damping, and the Bernstein waves should completely damp in the plasma periphery as they propagate toward a cyclotron harmonic resonance. The number of daughter waves is found to increase with rf power, and to increase as the incident wave's toroidal wavelength increases. The frequencies of the daughter wave are separated by the edge ion cyclotron frequency. Theoretical calculations of the threshold for this decay in uniform plasma indicate an extremely small value of incident power should be required to drive the instability. While such decays are commonly observed at lower harmonics in conventional ICRF heating scenarios, they usually do not involve the loss of significant wave power from the pump wave. On NSTX an estimate of the power loss can be found by calculating the minimum power required to support the edge ion heating (presumed to come from the decay Bernstein wave). This calculation indicates at least 20-30% of the incident rf power ends up as decay waves.

The advanced tokamak (AT) capability of the FIRE burning plasma experiment is examined with 0D systems analysis, equilibrium and ideal MHD stability, RF current drive analysis, and full discharge dynamic simulations. These analyses have identified the required parameters for attractive burning AT plasmas, and indicate that these are feasible within the engineering constraints of the device.

Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

Ion absorption of the high harmonic fast wave in a spherical torus is of critical importance to assessing the viability of the wave as a means of heating and driving current. Analysis of recent National Spherical Torus Experiment (NSTX) shots has revealed that under some conditions when neutral beam and radio-frequency (RF) power are injected into the plasma simultaneously, a fast ion population with energy above the beam injection energy is sustained by the wave. In agreement with modeling, these experiments find the RF-induced fast ion tail strength and neutron rate at lower B-fields to be less enhanced, likely due to a larger beta profile, which promotes greater off-axis absorption where the fast ion population is small. Ion loss codes find the increased loss fraction with decreased B insufficient to account for the changes in tail strength, providing further evidence that this is an RF interaction effect. Though greater ion absorption is predicted with lower k(sub)||, surprisingly little variation in the tail was observed, along with a small neutron rate enhancement with higher k(sub)||. Data from the neutral particle analyzer, neutron detectors, X-ray crystal spectrometer, and Thomson scattering is presented, along with results from the TRANSP transport analysis code, ray-tracing codes HPRT and CURRAY, full-wave code and AORSA, quasi-linear code CQL3D, and ion loss codes EIGOL and CONBEAM.

The two-dimensional radial versus poloidal structure and motion of edge turbulence in NSTX (National Spherical Torus Experiment) were measured by using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analyzed for Ohmic, L-mode (low-confinement mode) and H-mode (high-confinement mode) plasma conditions. Typical edge turbulence poloidal correlation lengths as measured using this technique are = 4 {+-} 1 cm and autocorrelation times are 40 {+-} 20 {micro}sec in all three regimes. The relative fluctuation level is typically smaller in H-mode than in L-mode, and transitions from H- to L-mode and can occur remarkably quickly (=30 {micro}sec). The two-dimensional images often show localized regions of strong light emission which move both poloidally and radially through the observed region at a typical speed of =10{sup 5} cm/sec, and sometimes show spatially coherent modes.

The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.

FIRE is a high-field, burning-plasma tokamak that is being studied as a possible option for future fusion research. Preliminary parameters for this machine are R/sub 0/=2 m, a=0.5 m, B/sub 0/=10 T, and I/sub p/=6 MA. Magnetic field coils are to be made of copper and pre-cooled with LN/sub 2/ before each shot. The flat-top pulse length desired is /spl ges/10 s. Ion cyclotron and lower hybrid rf systems will be used for heating and current drive. Present specifications call for 30 MW of ion cyclotron heating power, with 25 MW of lower hybrid power as an upgrade option.

The design of fast-wave current drive (FWCD) antennas combines the usual antenna considerations (e.g. the plasma-antenna interface, disruptions, high currents and voltages, and thermal loads) with new requirements for spectral shaping and phase control. The internal configuration of the antenna array has a profound effect on the spectrum and the ability to control phasing. The authors elaborate on these considerations, using a proof-of-principle (POP) experiment designed for the DIII-D tokamak. The extension of FWCD for machines such as the International Thermonuclear Experimental Reactor (ITER) will require combining ideas implemented in the POP experiment with reactor-relevant antenna concepts such as the folded waveguide. >

Ion absorption of the high harmonic fast wave in a spherical torus is of critical importance to assessing the viability of the wave as a means of heating and driving current. Analysis of recent NSTX shots has revealed that under some conditions when neutral beam and RF power are injected into the plasma simultaneously, a fast ion population with energy above the beam injection energy is sustained by the wave. In agreement with modeling, these experiments find the RF‐induced fast ion tail strength and neutron rate at lower B‐fields to be less enhanced, likely due to a larger β profile, which promotes greater off‐axis absorption where the fast ion population is small. Ion loss codes find the increased loss fraction with decreased B insufficient to account for the changes in tail strength, providing further evidence that this is an RF interaction effect. Though greater ion absorption is predicted with lower k∥, surprisingly little variation in the tail was observed, along with a small neutron rate enhancemen...