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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.

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.

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.

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 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.

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

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.

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.

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.

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. >

The design of the lower hybrid launcher is configured to couple 1.5 MW of steady state lower hybrid heating/lower hybrid current drive (LHH/LHCD) power at 3.7 GHz to the Tokamak Physics Experiment (TPX) plasma. The launcher utilizes a unique laminated, water-cooled construction to achieve a compact array of 128 waveguides at the grille surface and includes real-time position control for plasma coupling. The design features of this launcher are described with details of the principal components and the possible upgrade of the launcher to 3 MW. Results of steady state thermal analysis of the launcher grille septa are included.

The ICH system is required to deliver at least 20 MW of RF power to the BPX (Burning Plasma Experiment) plasma for pulse lengths of up to 15 s over a range of plasma operating conditions (e.g., average plasma densities from 0.5 to 4.5*10/sup 20/ m/sup -3/, toroidal fields from 5 to 8 T). The power must be delivered to the plasma over a range of frequencies corresponding to the different operating conditions. On the basis of calculations of the load resistance that the antennas will see because of the presence of the plasma, circuit analysis of the RF system indicates that the present ICH design should easily be able to deliver this amount of power to the plasma. Some research and development must be done to provide a reliable system in the BPX radiation environment. >

The design of an ICRF heating and current drive system was undertaken as part of the KSTAR (Korea Superconducting Tokamak Advanced Research) design study. ICRF fast wave heating and current drive scenarios are identified and modelled using the full wave code, TORIC which solves the finite Larmor radius wave equations in the ion cyclotron frequency range in an arbitrary axisymmetric toroidal geometry. The characteristics of the fast wave propagation/absorption, power partitions among plasma species, and the driven current/current drive efficiencies are analysed. The ICH/FWCD launcher is designed to have four current straps and each strap, using appropriate tuning, matching, and decoupling circuitry will be driven by an independent 2 MW RF power source. We present the design characteristics of the ICH/FWCD launcher, and the matching and decoupling system. Also, the feasibility of the antenna design to couple the required power to the plasma is demonstrated.

The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets are superconducting coils. The device is configured to be initially capable of 20 s pulse operation and then to be upgraded for operation up to 300 s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radiation shielding.

The lower hybrid (LH) system is part of an overall heating and current drive system for TPX. The LH system is required to provide off-axis current profile control, efficient bulk current drive at low plasma temperatures, and electron heating. The device and facility will provide 4 MW of generated LH/LHCD power with 3 MW of power to the plasma at 3.7 GHz through one horizontal port. This paper will describe the RF generator system that supplies RF power to the launcher. The LH launcher is described separately in a paper by Paul Goranson of ORNL.

The lower hybrid (LH) launcher configuration has been redesigned to integrate the functions of the vertical four-way power splitter and the front waveguide array (front array). This permits 256 waveguide channels to be fed by only 64 waveguides at the vacuum window interface. The resulting configuration is a more compact coupler, which incorporates the simplicity of a multijunction coupler while preserving the spectral flexibility of a conventional lower hybrid launcher. Other spin-offs of the redesign are reduction in thermal incompatibility between the front array and vacuum windows, improved maintainability, in situ vacuum window replacement, a reduced number of radio frequency (rf) connections, and a weight reduction of 7300 kg. There should be a significant cost reduction as well.

A 6-MW ion cyclotron (IC) system for the Tokamak Physics Experiment (TPX) is in the preliminary design phase. In conjunction with the 3-MW Lower Hybrid system and the 8-MW neutral beam system, the IC system will provide heating and current-drive capabilities to explore advanced tokamak physics and long-pulse (1000 s) operation. The IC launcher consists of six nickel-plated current straps arranged toroidally in pairs behind three water-cooled Faraday shields. The Faraday shields can be independently and remotely detached by cutting water lines at the back of the launcher and removing bolts at the front to free each shield. The antenna can be located at the +2 cm flux line and retracted 10 cm. Faraday shields are usually copper- or nickel-plated stainless steel or inconel. Titanium is the preferred material to minimize activation without greatly decreasing electrical resistivity and therefore increasing disruption loads. The IC antenna research and development programs have provided data that confirm the feasibility of B/sub 4/C-coated nickel-plated titanium alloy in the TPX environment.

The ion cyclotron (IC) and lower hybrid (LH) launchers for the Tokamak Physics Experiment (TPX) will be subjected to significant forces resulting from eddy currents induced in the launchers from the rapid decay of the plasma current during plasma disruption events. The launchers are being designed to withstand the forces resulting from a peak plasma current decay rate of 1.2 MA/ms and an average current decay rate of 0.5 MA/ms from an initial plasma current of 2 MA. The desire for highly reliable and low cost fabrication techniques prompted the comparison of pure copper and dispersion-hardened copper for the LH launcher. Its low nuclear activation rate, high electrical resistance, and high strength motivated the evaluation of Titanium as an IC antenna material.

The study of high harmonic fast wave (HHFW) heating and current drive is being conducted on the NSTX device to determine the physics of applying RF waves at high harmonics (~ 10-20) of the ion cyclotron frequency in this high β plasma regime and to extend the performance of the NSTX plasma. The magnetic field of this low aspect ratio device is lower (≤ 0.35 T for this work) than that for the typical moderate aspect ratio tokamak regime by about an order of magnitude and the plasma densities achieved are typically in the mid 10 19 m -3 range. Thus, the dielectric constant of the plasma, e ≡ ω pe 2 /ω ce 2 , is of order ~ 50-100 resulting in wave physics properties which favor electron heating by TTMP and Landau damping [1]. RF power is applied on NSTX at 30 MHz using an antenna array with 12 current straps aligned in the poloidal direction. The antenna can be phased to

ORNL and IPP-Garching are collaborating on fundamental studies of breakdown properties of ion cyclotron range of frequencies (ICRF) antennas in fusion experiments. The ORNL RF Breakdown Tester consists of a resonant 3/4-wavelength section of vacuum transmission line of compound impedance with a removable end piece for varying the geometry. Tests have been conducted with various gases to determine the pressure at which the breakdown voltage is affected by the gas fill. Other end pieces with different shapes will be used to determine geometrical effects of the breakdown, and an ultraviolet light source will be used to determine the effect of low-energy photons on the breakdown. The effects of materials will be explored by plating the end piece with nickel and silver. Other gases will be tested to determine their deleterious effects. Initial results the breakdown tester on ASDEX Upgrade will be reported.

Experimental operations are about to begin on the next generation of spherical torus (ST) devices-the National Spherical Torus eXperiment (NSTX) in the U.S. and the Mega-Amp Spherical Torus (MAST) in the U.K. The application of RF heating and current drive to these high beta, compact confinement devices is a challenging problem. The initial focus for NSTX had been on the High Harmonic Fast Wave (HHFW) regime. Although modeling of HHFW heating and current drive has been performed at ORNL, UCSD, MIT, and PPPL, there are few experiments in this frequency range. In conventional tokamaks, the DIII-D experiments at the 5th–7th cyclotron harmonic are the closest approach to the HHFW regime. In an ST, the only RF heating experiments to date have been performed at the 15th harmonic on the Current Drive eXperiment-Upgrade (CDX-U) at PPPL. General features of HHFW heating and current drive and the degree to which experimental confirmation of these features is available will be discussed.

This report describes the analysis and optimization of the proposed International Thermonuclear Experimental Reactor (ITER) Antenna Array for the ion cyclotron range of frequencies (ICRF). The objectives of this effort were to: (1) minimize the applied radiofrequency rf voltages occurring in vacuum by proper layout and shape of components, limit the component`s surface/volumes where the rf voltage is high; (2) study the effects of magnetic insulation, as applied to the current design; (3) provide electrical characteristics of the antenna for the development and analysis of tuning, arc detection/suppression, and systems for discriminating between arcs and edge-localized modes (ELMs); (4) maintain close interface with mechanical design.

Future fusion devices will require delivery of ion cyclotron heating and current drive power during plasma changes (e.g., L-H transition, ELMs). The use of a passive circuit (``ELM dump``) to protect the RF sources during transients has been demonstrated on DIII-D, and the results are applied to the ITER ion cyclotron system in this analysis. In addition, the use of frequency shifting to compensate for plasma load changes is illustrated for a possible ITER tuning and matching system.

Tokamak operation using advanced tokamak modes with steady-state current drive may provide better performance, thereby improving the prospects of power-producing fusion reactors. Scenarios for such operation in TPX and in an ITER-like scale-up of TPX are modeled. Results indicate that the heating and current drive power and feedback control requirements may be achievable.

This report discusses the following topics: The TST program; the TST physics basis; the TST auxiliary HC the test divertors; the TST device; and ancillary systems.