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

The critical issue of divertor configuration for heat and particle flux control in a conceptual ARIES compact stellarator (CS) reactor is addressed. The goal is to determine a divertor location and...

Alpha loss is an important issue in stellarators, where the inherent non-axisymmetry of the magnetic geometry causes an appreciable magnetic field ripple along the flux surfaces inside the plasma. This results in a non-negligible fraction of alpha-particles being promptly lost from the plasma and hitting the plasma facing components (PFC) at energies close to their born value of 3.5 MeV. The PFC armor must not only accommodate the heat load from the alpha-particle flux but it must also accommodate the He implantation resulting from the high-energy alpha fluxes while providing the required lifetime. An effort is underway as part of the ARIES-CS study to characterize the alpha-particle loss for a compact stellarator and to explore engineering solutions to accommodate the impact of these alphas on the PFC’s. This paper summarizes the initial results from this effort.

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.

An advanced tokamak plasma configuration is developed based on equilibrium, ideal MHD stability, bootstrap current analysis, vertical stability and control, and poloidal field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current drive profiles from ray tracing calculations in combination with optimized pressure profiles, βN values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower βN of 6.0. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field from those found in a previous study [S.C. Jardin, C.E. Kessel, C.G. Bathke, D.A. Ehst, T.K. Mau, F. Najmabadi, T.W. Petrie, the ARIES Team, Physics basis for a reversed shear tokamak power plant, Fusion Eng. Design 38 (1997) 27].

The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A ≡ R / a = 4.0 , an elongation and triangularity of κ = 2.20 , δ = 0.90 (evaluated at the separatrix surface), a toroidal beta of β = 9.1 % (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of β N ≡ 100 × β / ( I P ( M A ) / a ( m ) B ( T ) ) = 5.4 . These beta values are chosen to be 10% below the ideal MHD stability limit. The bootstrap-current fraction is f BS ≡ I BS / I P = 0.91 . This leads to a design with total plasma current I P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current drive system consists of ICRF/FW for on-axis current drive and a Lower Hybrid system for off-axis. Transport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (βN = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κx = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher βN. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then

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.

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...

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.

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 ...

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/.

In inertial fusion energy (IFE) power plant designs, the fuel is a spherical layer of frozen DT contained in a target that is injected at high velocity into the reaction chamber. For direct drive, typically laser beams converge at the centre of the chamber (CC) to compress and heat the target to fusion conditions. To obtain the maximum energy yield from the fusion reaction, the frozen DT layer must be at about 18.5 K and the target must maintain a high degree of spherical symmetry and surface smoothness when it reaches the CC. During its transit in the chamber the cryogenic target is heated by radiation from the hot chamber wall. The target is also heated by convection as it passes through the rarefied fill-gas used to control chamber wall damage by x-rays and debris from the target explosion. This article addresses the temperature limits at the target surface beyond which target uniformity may be damaged. It concentrates on direct drive targets because fuel warm up during injection is not currently thought to be an issue for present indirect drive designs and chamber concepts. Detailed results of parametric radiative and convective heating calculations are presented for direct-drive targets during injection into a dry-wall reaction chamber. The baseline approach to target survival utilizes highly reflective targets along with a substantially lower chamber wall temperature and fill-gas pressure than previously assumed. Recently developed high-Z material coatings with high heat reflectivity are discussed and characterized. The article also presents alternate target protection methods that could be developed if targets with inherent survival features cannot be obtained within a reasonable time span.

In recent experiments with combined fast wave current drive (FWCD) and deuterium neutral beam injection on the DIII-D tokamak [Luxon et al., Fusion Technol. 8, 441 (1985)], an enhanced fusion reactivity and fast ion energy content have been observed in the presence of FWCD, with a concomitant low FWCD efficiency [Petty et al., Radio Frequency Power in Plasmas (AIP, New York, 1997), p. 225]. In this paper, we investigate whether high-harmonic ion cyclotron damping could be responsible for the low FWCD efficiency in these experiments, since a number of high-harmonic hydrogen and deuterium cyclotron resonance layers existed in the plasma. The main analysis tool is the ion cyclotron range of frequencies code PION [Eriksson et al., Nucl. Fusion 33, 1037 (1993)], modified to allow multiple frequencies simultaneously as was done in the DIII-D experiments. According to the PION modeling, high harmonic damping of fast wave power can give rise to enhanced fusion reactivity and fast ion energy content, which is cons...

Current driven by the fast Alfven wave is measured at the fourth and eighth harmonics of the deuterium ion cyclotron frequency in identical plasmas on the DIII-D tokamak. In non-sawtoothing discharges with neutral beam injection heating, the radial profile of the fast wave current drive (FWCD) is determined by the response of the loop voltage profile to co- and counter-antenna phasings. The dimensionless current drive efficiency is a factor of two greater for the eighth harmonic case compared to the fourth harmonic case. Modelling of the fast wave absorption using a ray tracing code shows that the decrease in FWCD efficiency for the latter situation can be explained by high harmonic damping of the fast waves on energetic ions.

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 five ARIES designs, which correspond to five different tokamak operating modes, are reviewed and compared. Physics figures of merit are introduced that quantify the major parameters of a tokamak design in a physics operating space. The five operating modes are compared to one another and to the existing tokamak data base in terms of these physics parameters. While the steady-state first stability design [ARIES-I] and the pulsed first stability design [PULSAR] are closest, no design has yet been completely prototyped in existing tokamaks.

Current driven by fast Alfven waves is measured in H mode and VH mode plasmas on the DIII-D tokamak for the first time. Analysis of the poloidal flux evolution shows that the fast wave current drive profile is centrally peaked but sometimes broader than theoretically expected. Although the measured current drive efficiency is in agreement with theory for plasmas with infrequent ELMs, the current drive efficiency is an order of magnitude too low for plasmas with rapid ELMs. Power modulation experiments show that the reduction in current drive with increasing ELM frequency is due to a reduction in the fraction of centrally absorbed fast wave power. The absorption and current drive are weakest when the electron density outside the plasma separatrix is raised above the fast wave cut-off density by the ELMs, possibly allowing an edge loss mechanism to dissipate the fast wave power since the cut-off density is a barrier for fast waves leaving the plasma.

Progress in theory and in tokamak experiments leads to questions of the optimal development path for commercial tokamak power plants. The economic prospects of future designs are compared for several tokamak operating modes: (high poloidal beta) first stability, second stability and reverse shear. Using a simplified economic model and selecting uniform engineering performance parameters, this comparison emphasizes the different physics characteristics - stability and non-inductive current drive - of the various equilibria. The reverse shear mode of operation is shown to offer the lowest cost of electricity for future power plants.

ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems.

The reversed shear plasma configuration is examined as the basis for a tokamak fusion power plant. Analysis of plasma equilibrium and ideal MHD stability, bootstrap current and current drive, plasma vertical stability and position control, divertor physics and plasma power balance are used to determine the operating point parameters that maximize fusion power density and minimize the recirculating power fraction. The final plasma configuration for the ARIES-RS power plant obtains b of 4.96%, plasma driven current fraction of 91%, plasma current of 11.3 MA, toroidal field of 8.0 T and major and minor radius of 5.5 and 1.4 m. The current drive system utilizes fast wave, lower hybrid and high frequency fast wave current drive to obtain maximum current profile flexibility, requiring 5 80 MW of power. A divertor solution is found which employs neon impurity injection to enhance the radiation in the scrape-off layer (SOL) and divertor and results in a combined particle and heat load in the divertor of5 6M W m 2 . The plasma is driven with a Q of 25 and is at a thermally stable operating point. The plasma is assumed to be in an ELMy H-mode, with low amplitude and high frequency ELMs. © 1997 Elsevier Science S.A. All rights reserved.

The ARIES-RS tokamak is a conceptual, D‐T-burning 1000 MWe power plant. As with earlier ARIES design studies, the final design of ARIES-RS was obtained in a self-consistent manner using the best available physics and engineering models. Detailed analyses of individual systems together with system interfaces and interactions were incorporated into the ARIES systems code in order to assure self-consistency and to optimize towards the lowest cost system. The ARIES-RS design operates with a reversed-shear plasma and employs a moderate aspect ratio (A4.0). The plasma current is relatively low (Ip11.32 MA) and bootstrap current fraction is high ( fBC 0.88). Consequently, the auxiliary power required for RF current drive is relatively low ( 80 MW). At the same time, the average

The modeling steps needed to create dynamically based automated control of tokamak plasmas are traced. This involves integrating models of current/magnetic-flux dynamics; plasma transport; plasma geometry; and source terms, such as lower hybrid, fast wave, and pellet-fueling deposition. Perturbative analysis of these models then yields the linear response of the tokamak to changes in coil voltages, applied radio-frequency power, or pellet-firing frequency. Comparison of the linear response models to nonlinear numerical calculations reveals that the plasma position and shape modeling will require future refinements.

The magnetic field of a low aspect ratio tokamak (LART) deviates rather significantly from the approximately 1/R dependence of conventional tokamaks. The parameterization of current drive efficiency on the basis of conventional large aspect ratio tokamaks becomes inaccurate. The physics of high harmonic ion cyclotron fast wave current drive (FWCD) and low frequency FWCD in LARTs is examined using exact evaluations of the current drive efficiencies from the adjoint technique. The results indicate that the high harmonic ion cyclotron current drive cannot penetrate to the plasma core at the high beta values typical of LART reactors, and hence this scheme may only be applicable for off-axis steady state current drive, while the low frequency current drive penetrates well into the plasma core and is appropriate for on-axis current profile control. The implications for advanced scenarios of LART are discussed

Results for the first simulated comprehensive feedback control study for a tokamak operating in the fusion regime are presented. A standard Burning Plasma Experiment (BPX) design is the simulated reactor, but the results apply to any tokamak. Feedback gains are derived for specific classes ofdynamic models and control objectives using model-based optimal control. An integrated control approach treats both kinetic and electromagnetic parameters and radial profiles. The control actuators include poloidal field coils, fast-wave and lower-hybrid current drive and heating sources, and pellet fuel injectors. Results show that the strongly coupled plasma parameters provide unintended secondary responses to controller inputs. In particular, attempts to modify the q-profile greatly affect the temperature and density profiles when the transport model incorporates International Thermonuclear Experimental Reactor(ITER) scaling. The hot, highly conductive plasma and poor source penetration in the nominal BPX discharge make the central q-values difficult to regulate. Fusion events also complicate the control efforts. Further, simple plasma circuit models are inadequate to account for a significantly evolving current profile. Proper understanding and use of integrated, model-based feedback control will avoid these pitfalls.

A weakly relativistic theory of current drive by parallel acceleration of electrons in toroidal geometry is presented. Analytic expressions of efficiency are obtained for superthermal and subthermal electrons. By using a few adjustable parameters for intermediate energies, an accurate formula for the efficiency is obtained by fitting numerical results. This includes trapping effects and is valid for all frequencies up to the lower hybrid range of frequencies. It is found that for low‐frequency fast‐wave current drive, the relativistic and nonrelativistic efficiencies agree to within 15% up to 500 keV, although the relativistic corrections to power absorption and current density can be much larger separately. Relativistic effects on the efficiency become prominent at lower energies for Landau damping.

Steady-state plasma operating scenarios are designed for three versions of the ARIES reactor, using non-inductive current drive techniques that have an established database. R.f. waves, including fast and lower hybrid waves, are the reference drivers for the D–T burning ARIES-1 and ARIES-II/IV, while neutral beam injection is employed for ARIES-III which burns D–3He. Plasma equilibria with a high bootstrap-current component have been used, in order to minimize the recirculating power fraction and cost of electricity. To maintain plasma stability, the driven current profile has been aligned with that of equilibrium by proper choices of the plasma profiles and power launch parameters. Except for ARIES-III, the current-drive power requirements and the relevant technology developments are found to be quite reasonable. The wave-power spectrum and launch requirements are also considered achievable with a modest development effort. Issues such as an improved database for fast-wave current drive, lower-hybrid power coupling to the plasma edge, profile control in the plasma core, and access to the design point of operation remain to be addressed.

Fast waves in the ion cyclotron range of frequencies are proposed as the reference current driver for the ARIES-I conceptual tokamak power reactor. An RF power of 100 MW at 141 MHz is launched from above the plasma outboard midplane to drive the required 3.3 MA seed current in the core. Folded waveguides are used as wave launchers because of their observed higher power-handling capability compared to loops, and compact and robust structure. The wave guide will be made of silicon-carbide matrix composite material embedded with copper near the surface, in order to take advantage of its favorable safety and environmental features. The entire coupler system is compact, consisting of two poloidally stacked 12-waveguide toroidal arrays with a directivity of ≤ 95% and a coupling efficiency of ∼ 97%. The RF current-drive system uses primarily technologies that require a reasonably modest development program to reach its projected efficiency of 71% in 30–40 years.

The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currentl...

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

Fast waves launched from folded waveguides are proposed as the reference current drive scenario for the ARIES-I tokamak reactor. An RF power of 100 MW at 141 MHz is launched from above the plasma m...

Divertor heat load distributions due to thermal and alpha particles have been assessed in an NCSX-based compact stellarator reactor. A divertor plate system is envisaged, with 4 plates per field period and covering 7% of the plasma surface area. The field-line tracing technique is employed; for thermal flux, the conventional approach is used, while for alphas, their characteristic exit pattern from the plasma and subsequent gyro- orbits are approximated. For the ARIES-CS reference design point (R=7.75 m, A=4.5, B=5.7 T, beta=6.4% and Pnet=1000 MW), combined peak heat loads in the 5 -18 MW/m2 range on the plates have been obtained, assuming a 75% radiation fraction both in the core and at the edge, and a 5% alpha loss fraction. The alpha heat flux could be a dominant determining factor. Further optimization study is warranted to lower all peak heat loads to satisfy the accepted limit of les10 MW/m2.

Divertor heat load evaluation studies are being carried out for an example compact stellarator reactor configuration modeled after the National Compact Stellarator Experiment (NCSX). Using the field line tracing technique, an initial divertor plate geometry has been obtained, with one plate per field period, which covers 15% of the plasma surface area. A plate heat load peaking factor of 14 is acceptable with reasonable core and SOL radiation fractions in order to satisfy the engineering constraint of 10 MW/m peak heat load. Assessment of heat flux due to ejected energetic alpha particles is equally important for their contribution to the heat load. A particle gyro-orbit code follows these particles in the scrape-off layer (SOL) until they hit the divertor plates and the first wall, while the distribution of lost particles on the plasma surface can be calculated with a Monte Carlo guiding center code with collisions

Integrated scenario simulations are done for NSTX [National Spherical Torus Experiment] that address four primary milestones for developing advanced ST configurations: high {beta} and high {beta}{sub N} inductive discharges to study all aspects of ST physics in the high-beta regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current-drive techniques; non-inductively sustained discharges at high {beta} for flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX; and non-solenoidal start-up and plasma current ramp-up. The simulations done here use the Tokamak Simulation Code (TSC) and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral-beam (NB) deposition profile, and other characteristics. CURRAY is used to calculate the High Harmonic Fast Wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD [current drive] deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal-MHD stability is done with JSOLVER, BALMSC, and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with {beta} {approx} 40% at {beta}{sub N}'s of 7.7-9, I{sub P} = 1.0 MA, and B{sub T} = 0.35 T. The plasma is 100% non-inductive and has a flattop of 4 skin times. The resulting global energy confinement corresponds to a multiplier of H{sub 98(y,2)} = 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control, and early heating/H-mode transition for producing and optimizing these plasma configurations.

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.

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

Fast-wave current drive is selected as the primary scenario for steady-state operation of the Advanced Reactor Innovation and Evaluation Study-I (ARIES-I) tokamak reactor. With a projected efficiency of approximately=0.3-0.4, the current drive method uses low-cost, off-the-shelf technology, making it attractive compared to neutral beam current drive. At a frequency of 158 MHz and a wave spectrum centered at 1.6, it is found that as much as 160 MW of power may be required to drive the 4.7-MA seed current on the ARIES-I. To generate the desired current profile, the wave power is launched from above the equatorial plane on the outboard edge of the plasma. A toroidal array of 12 folded waveguides forms the basic launcher module which transmits the desired spectrum with a 95% directivity. Efficiency of coupling to the high-density scrape-off layer is expected to be high (>90%). The overall launcher system consists of four modules delivering 192 MW through 1.6% of the first-wall area. Each waveguide is envisaged to be a closed rectangular box made of SiC/SiC fiber composites, with the conducting surfaces formed by a copper coating on the inside of the structure. With the klystrode as the transmitter, the overall RF system efficiency can be as high as 72%. A viable alternative system based on neutral beam current drive and RFQ (radio frequency quadrupole technology has also been designed. >

High-efficiency energy conversion schemes for a D-/sup 3/He tokamak reactor are investigated. The loss channels for a D-/sup 3/He plasma are through charged particles, neutrons, and radiation (synchrotron and bremsstrahlung), with only a few percent of the energy loss in neutrons. In contrast to D-T fusion reactors, this gives high leverage to energy-conversion methods which apply to charged particles or radiation. The ideas investigated focus on efficiently converting such energy to electricity or on converting thermal energy in ways which make effective use of fusion-reactor characteristics such as high magnetic fields. The energy-conversion methods investigated are given. Whether a concept was chosen for further study is indicated, as is the applicability of the concept to the three tokamak versions under consideration for the Advanced Reactor Innovation and Evaluation Study-III (ARIES-III): a high-field reactor (HFR), a second stability reactor (SSR), and a spherical torus (ST). Options not selected for further pursuit within the ARIES project are described briefly, and those selected are discussed more extensively. Selection criteria include cost, efficiency, technical feasibility, and how well a concept made use of fusion-specific features. >

The Advanced Reactor Innovation and Evaluation Study (ARIES) is a community effort to develop several visions of the tokamak as a fusion power reactor. The aims are to determine its potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. The authors focus on the ARIES-1 design. Parametric systems studies show that the optimum first stability tokamak has relatively low plasma current ( approximately 12 MA), high plasma aspect ratio ( approximately 4-6), and high magnetic field ( approximately 24 T at the coil). ARIES-I is a 1000-MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m/sup 2/, and a mass power density of about 90 kWe/tonne. The ARIES-I reactor operates at steady state using ICRF (ion-cyclotron range of frequency) fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The ARIES-I blanket is cooled by He and consists of SiC-composite structural material, Li/sub 4/SiO/sub 4/ solid breeder, and Be neutron multiplier, all chosen for their low-activation and low-decay after-heat in order to enhance the safety and environmental features of the design. The ARIES-I design has a competitive cost of electricity and superior safety and environmental features. >

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...

Phased‐array experiments were performed with the 30‐MHz, 12‐element HHFW antenna array on NSTX for rf powers in the 2–4.5 MW range. Relative phase shift of the array was scanned from π/6 to π/2 to create kz spectral peaks between 3 and 7.6 m−1; it was also possible to switch from π/2 to π/4 phasing during a shot. Typical plasma parameters were Ip ∼ 0.5 MA, BT ∼ 0.45 T, and ne(0) ∼ 1−3×1019 m−3, and Te(0) ∼ 1−3 keV. Electron heating was observed even at the highest wave phase velocities, reaching 1.3 keV with 2.2 MW at −π/6 phasing and 2.7 keV with 2.9 MW at −π/3. Lower plasma loading and higher central heating efficiency are observed for counter‐CD phasing. MSE measurement of the current profile is not yet available; therefore, the rf‐driven portion of the plasma current is calculated from the difference in loop voltage between co‐ and counter‐CD phasing. The normalized current drive efficiency for kz = 7.6 m−1 is γ∼ 0.035×1019 A•W−1•m−2 for both D and He plasmas.