L. Grisham

A component test facility based on the spherical tokamak

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 (wall loading WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year (neutron fluence >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~3, Ip ~ 12 MA, BT ~ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

Moderate energy ions for high energy density physics experiments

This paper gives the results of a preliminary exploration of whether moderate energy ions (≈0.3–3u2002MeV/amu) could be useful as modest-cost drivers for high energy density physics experiments. It is found that if the target thickness is chosen so that the ion beam enters and then leaves the target in the vicinity of the peak of the dE/dX (stopping power) curve, high uniformity of energy deposition may be achievable while also maximizing the amount of energy per beam particle deposited within the target.

Operation and Development of the 500-keV Negative-Ion-Based Neutral Beam Injection System for JT-60U

The 500-keV negative-ion based neutral beam injector for JT-60U started operation in 1996. The beam power has been increased gradually through optimizing operation parameters of the ion sources and conquering many troubles in the ion source and power supplies caused by a high voltage break-down in the accelerator. However, some issues remain to be solved concerning the ion source for increasing further the beam power and the beam energy. The most serious issue of them is non-uniformity of source plasma in the arc chamber. Various countermeasures have been implemented to improve the non-uniformity. Some of those countermeasures have been found to be partially effective in reducing the non-uniformity of the source plasma, and as the result the ion source, so far, has accelerated negative-ion beams of 17.4A at 403keV with deuterium and 20A at 360keV with hydrogen against the goal of 22A at 500keV. The neutral beam injection power into the plasma has reached 5.8MW at 400keV with deuterium. Further efforts to reach the target of 10MW at 500keV have been continued.

Overview of KSTAR initial operation

Since the successful first plasma generation in the middle of 2008, three experimental campaigns were successfully made for the KSTAR device, accompanied with a necessary upgrade in the power supply, heating, wall-conditioning and diagnostic systems. KSTAR was operated with the toroidal magnetic field up to 3.6u2009T and the circular and shaped plasmas with current up to 700u2009kA and pulse length of 7u2009s, have been achieved with limited capacity of PF magnet power supplies.The mission of the KSTAR experimental program is to achieve steady-state operations with high performance plasmas relevant to ITER and future reactors. The first phase (2008–2012) of operation of KSTAR is dedicated to the development of operational capabilities for a super-conducting device with relatively short pulse. Development of start-up scenario for a super-conducting tokamak and the understanding of magnetic field errors on start-up are one of the important issues to be resolved. Some specific operation techniques for a super-conducting device are also developed and tested. The second harmonic pre-ionization with 84 and 110u2009GHz gyrotrons is an example. Various parameters have been scanned to optimize the pre-ionization. Another example is the ICRF wall conditioning (ICWC), which was routinely applied during the shot to shot interval.The plasma operation window has been extended in terms of plasma beta and stability boundary. The achievement of high confinement mode was made in the last campaign with the first neutral beam injector and good wall conditioning. Plasma control has been applied in shape and position control and now a preliminary kinetic control scheme is being applied including plasma current and density. Advanced control schemes will be developed and tested in future operations including active profiles, heating and current drives and control coil-driven magnetic perturbation.

Initial physics results from the National Spherical Torus Experiment

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.

Improvement of beam performance in the negative-ion based NBI system for JT-60U

The injection performance of the negative-ion based NBI (N-NBI) system for JT-60U has been improved by correcting beamlet deflection and improving spatial uniformity of negative ion production. Beamlet deflection at the peripheral region of the grid segment due to the distorted electric field at the bottom of the extractor has been observed. This was corrected by modifying the surface geometry at the extractor to form a flat electric field. Moreover, beamlet deflection due to beamlet–beamlet repulsion caused by space charge was also compensated for by extruding the edge of the bottom extractor. This resulted in a reduction of the heat loading on the NBI port limiter. As a result of the improvement above, continuous injection of a 2.6 MW H0 beam at 355 keV has been achieved for 10 s. Thus, long pulse injection up to the nominal pulse duration of JT-60U was demonstrated. This has opened up the prospect of long pulse operation of the negative-ion based NBI system for a steady-state tokamak reactor. So far, a maximum injection power of 5.8 MW at 400 keV, with a deuterium beam, and 6.2 MW at 381 keV, with a hydrogen beam, have been achieved in the JT-60U N-NBI. Uniformity of negative ion production was improved by tuning the filament emission current so as to direct more arc power into the region where less negative ion current was extracted.

Notched velocity profiles and the radial electric field in high ion temperature plasmas in the Tokamak Fusion Test Reactor

A large “notch,” or non-monotonic feature, appears in measured toroidal velocity profiles of the carbon impurity in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], centered near the radius of strongest ion temperature gradient. This is explained as a consequence of radial momentum transport dominated by anomalous diffusion together with parallel heat friction on the impurity ions arising from the hydrogenic neoclassical parallel heat flow. The toroidal velocity profile of the hydrogenic species is predicted to be monotonic, from measurements of the impurity toroidal velocity, consistent with the anomalous radial diffusion of toroidal momentum. This supports a neoclassical calculation of the radial electric field for near-balanced beam injection. In supershot plasmas [Phys. Rev. Lett. 58, 1004 (1987)], a well structure in the radial electric field profile is found in the enhanced confinement region. An associated shear layer separates the core, where the local confine...

Achievement and improvement of the JT-60U negative ion source for JT-60 Super Advanced (invited)

Developments of the large negative ion source have been progressed in the high-energy, high-power, and long-pulse neutral beam injector for JT-60 Super Advanced. Countermeasures have been studied and tested for critical issues of grid heat load and voltage holding capability. As for the heat load of the acceleration grids, direct interception of D- ions was reduced by adjusting the beamlet steering. As a result, the heat load was reduced below an allowable level for long-pulse injections. As for the voltage holding capability, local electric field was mitigated by tuning gap lengths between large-area acceleration grids in the accelerator. As a result, the voltage holding capability was improved up to the rated value of 500 kV. To investigate the voltage holding capability during beam acceleration, the beam acceleration test is ongoing with new extended gap.

Present status of the negative ion based NBI system for long pulse operation on JT-60U

The 500 keV negative-ion based neutral beam injector for JT-60U started operations in 1996. The availability of the negative ion based neutral beam injection system has been improved gradually by modifying the ion source and optimizing its operation parameters. Recently, the extension of the pulse duration up to 30 s has been intended to study quasi-steady state plasma on JT-60U. The most serious issue is to reduce the heat load on the grids for long pulse operation. Two modifications have been proposed to reduce the heat load. One is to suppress the spread of beamlet-bundle which may be caused by beamlet–beamlet interaction in the multi-aperture grid due to the space charge force. Indeed, the investigation of the beam deflection, which was measured by the infrared camera on the target plate set 3.5 m away from the grid, indicates that the spread of beamlet-bundle is in proportion to the current density. Field-shaping plates were attached on the extraction grid to modify the local electric field. The plate thickness was optimized to steer the beamlet deflection. The other is to reduce the stripping loss, where the electron of the negative ion beam is stripped and accelerated in the accelerator and then collides with the grids. The ion source was modified to reduce the pressure in the accelerator column to suppress the beam-ion stripping loss. To date, long pulse injection of 19 s of 1.5–1.6 MW at a high energy beam of 360 keV, 9–10 A for D− has been obtained by one ion source with these modifications.

Status and plans for TFTR

AbstractRecent research on TFTR has emphasized optimization of performance in deuterium plasmas, transport studies and studies of energetic ion and fusion product physics in preparation for the D-T experiments that will commence in July of 1993. TFTR has achieved full hardware design parameters, and the best TFTR discharges in deuterium are projected to QDT of 0.3 to 0.5.The physics phenomena that will be studied during the D-T phase will include: tritium particle confinement and fueling, ICRF heating with tritium, species scaling with tritium, collective alpha-particle instabilities, alpha heating of the plasma and helium ash buildup. It is important for the fusion program that these physics issues be addressed to identify regimes of benign alpha behavior, and to develop techniques to actively stabilize or control instabilities driver by collective alpha effects.