Inside Launch Research Articles

Pellet delivery and survivability through curved guide tubes for fusion fueling and its implications for ITER

Injection of solid hydrogen pellets from the magnetic high-field side will be the primary technique for depositing fuel particles into the core of International Thermonuclear Experimental Reactor (ITER) burning plasmas. This injection scheme will require the use of curved guide tubes to route the pellets from the acceleration devices to the inside wall launch locations. Experimental pellet studies with curved guide tubes have been carried out for several years at the Oak Ridge National Laboratory (ORNL), including mock-up tests of guide tube installations for several fusion experiments. In general, the tightest bend radius of the guide tube is the key parameter in limiting intact pellet speed, and for inside launch the pellet speed is typically limited to a few hundreds of meters per second for reliable delivery of intact pellets. Recently, an experimental mock-up of the proposed ITER curved guide tube for inside wall pellet launch was set up in the lab and tested with nominal 5.3-mm D 2 pellets. For this mock-up, the pellet speed had to be limited to ≈300 m/s for reliable delivery of intact pellets. Also, microwave cavity mass detectors located upstream and downstream of the test tube indicated that ≈10% of the pellet mass was lost in the guide tube at 300 m/s. The test results from the previous mock-ups will be summarized in the paper, and the new data from the ITER mock-up will be presented and discussed.

ORNL mock-up tests of inside launch pellet injection on JET and LHD

In experiments on ASDEX-Upgrade and DIII-D tokamaks, the injection of D 2 pellets from the magnetic high-field side of the plasma resulted in deeper pellet penetration and improved fueling efficiency. Based on those successful experiments, fusion researchers at the Joint European Torus and the Large Helical Device decided to implement inside launch pellet injection. These injection schemes require the use of curved guide tubes to route the pellets from the acceleration devices to the inside launch locations, and the pellets are subjected to stresses from centrifugal and impact forces in traversing the tubes. Before the installations on the large experimental fusion devices, mock-ups of the guide tubes were constructed and tested at the Oak Ridge National Laboratory to determine the pellet speed limit for reliable operation without pellet fracturing. In laboratory testing of the mock-ups, it was found that the pellet speed had to be limited to a few hundreds of meters per second for intact pellets. In this paper, the test equipment and experimental results are described.

Physics fundamentals for ITER

The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an `empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should be forthcoming soon. Recent results on JET and TFTR have confirmed qualitative understanding of alpha particle driven toroidal Alfvén eigenmodes (TAEs). Present predictions for TAE effects in ITER are favourable, but require further work. The large stored energies in ITER have focused attention on disruption physics. Databases for thermal and current quenches, vertical displacement events (VDEs) and halo currents have enabled thermomechanical design. Some questions remain open as to the production, confinement and localization of runaway electrons in potentially unstable plasmas and mitigation strategies have been proposed. Other crucial ITER needs such as diagnostics, control and heating appear to have acceptable solutions. All this rich physics requires experimental validation by a reactor-scale plasma and care has been taken to provide sufficient flexibility for ITER to cover a wide range of scenarios.

High-Field-Side Pellet Injection Technology

High-speed injection of pellets, composed of frozen hydrogen isotopes and multimillimeter in size, is commonly used for core fueling of magnetically confined plasmas for controlled thermonuclear fusion research. Straight guide tubes have typically been used to transport/deliver pellets from the acceleration device to the outside, or magnetic low-field side, of the torus/plasma (distance of −5 to 10 m for most installations). Recently, alternative pellet injection schemes have been used in plasma fueling experiments, including inside launch from the magnetic high-field side on ASDEX-U and top launch (vertically downward) on Tore Supra and DIII-D. These schemes require the use of curved guide tubes in which the pellets are subjected to stresses from centrifugal and impact forces. Thus, with curved guide tubes the speed at which intact pellets can be delivered reliably to the plasma is limited. In impact experiments on flat plates, it was found that deuterium (D2) pellets can survive single collisions at normal velocities in the range 20 to 35 m/s. Several series of tests with various curved guide tube configurations have been carried out, showing that intact pellets can be reliably delivered at speeds of several hundreds of meters per second. The experimental data are summarized and discussed. Also, a model is under development at Tore Supra for predicting these phenomena, and preliminary comparisons with the data are discussed.

Development of inside launch reflectometer systems on the DIII-D tokamaka)

Inside launch (high field side) reflectometry is necessary for routine core access on DIII-D and is a very attractive option for ITER and TPX. For high temperature reactor relevant plasmas, relativistic corrections for the left-hand cutoff (inside launched) are relatively small, whereas these substantially affect the location of the normal right hand cutoff. On DIII-D, core access using the right hand cutoff is restricted to low density plasmas, while O-mode polarization is incompatible with flat H-mode density profiles. Routine core access can, however, be obtained using the left-hand cutoff and inside launch. On ITER and TPX, outside launch X-mode systems need to operate at high frequency, ∼100–250 GHz, such that the left-hand cutoff, which is at much lower frequency, ∼0–75 GHz, is an attractive option for high density operation. In addition, inside launch reflectometry can be used to investigate plasma asymmetries; on DIII-D an inboard/outboard asymmetry in turbulence response is observed at the L-H transition. An upgraded inside launch reflectometer system was recently installed on DIII-D, specifically to investigate the use of the left hand X-mode cutoff and inside launch for profile and turbulence measurements in reactor relevant plasmas. This system will also provide routine core access for reflectometry on DIII-D.

Development of inside launch reflectometer systems on the DIII-D tokamak (abstract)a)

Inside launch (high field side) reflectometry is necessary for routine core access on DIII-D and is a very attractive option for ITER and TPX. For high temperature reactor relevant plasmas, relativistic corrections for the left-hand cutoff (inside launched) are relatively small, whereas these substantially affect the location of the normal right hand cutoff. On DIII-D, core access using the right hand cutoff is restricted to low density plasmas, while O-mode polarization is incompatible with flat H-mode density profiles. Routine core access can, however, be obtained using the left-hand cutoff and inside launch. On ITER and TPX, outside launch X-mode systems need to operate at high frequency, ∼100–250 GHz, such that the left-hand cutoff, which is at much lower frequency, ∼0–75 GHz, is an attractive option for high density operation. In addition, inside launch reflectometry can be used to investigate plasma asymmetries; on DIII-D an inboard/outboard asymmetry in turbulence response is observed at the L-H transition. An upgraded inside launch reflectometer system was recently installed on DIII-D, specifically to investigate the use of the left hand X-mode cutoff and inside launch for profile and turbulence measurements in reactor relevant plasmas. This system will also provide routine core access for reflectometry on DIII-D.

2-D FIR interferometry and polarimetry on TPX

A multiple fan-beam approach to 2-D FIR interferometry and polarimetry is under investigation for use on next generation tokamaks such as TPX. The approach utilizes a small number (3–5) of inside launch fan beams, and a finite number (5–15) of receive horns positioned within the vacuum vessel on the outboard side of the plasma. Positioning of the horns is achieved with minimal penetration of internal structures, within the design constraints dictated by a highly restricted viewing access. The choice of operating frequency is discussed considering both physics and design issues.

New reflectometer systems for the DIII-D tokamak (abstract)

During a machine vent in December 1991, two new reflectometer systems were successfully installed and tested on the DIII-D tokamak. The first is an X-mode broadband system primarily intended for density profile measurements, utilizing BWO sources and covering Q and V frequency bands (33–50 and 50–75 GHz). The second system is an adaptation of a pre-existing inside launch (high field side) ECRH waveguide to provide an inside launch reflectometer capability at the same frequencies and polarization as an outside launch fixed frequency O-mode system. The new systems will have a dual role in both directly supporting the DIII-D physics program, and also acting as flexible and adaptable test beds for the development of reactor relevant reflectometer systems, such as required for ITER. Specific examples of planned measurements include investigation of possible in/out plasma asymmetries at the L–H transition and ELMs, and demonstration of routine and reliable density profile measurements. It is expected that preliminary data from the inside launch system will be available by the time of the conference. This work is supported by the U. S. Department of Energy under Grant No. DE-FG03-86-ER53225 and General Atomics subcontract SC120536 under DOE Contract No. DE-AC03-89ER51114.

Interpretation of electron cyclotron heating results in overdense plasma in Doublet III

The electron cyclotron heating (ECH) experiment on the Doublet III tokamak with inside oblique launch of the extraordinary mode has shown good heating even at very high densities, up to the operational density limit. According to ray tracing calculations for the high density discharges, the ECH waves are reflected at the periphery of the plasma around r/a ∼ 0.9, and bulk plasma heating is not expected theoretically. But effective good heating has been observed in the increases of both the central electron temperature measured by Thomson scattering and the total energy measured magnetically. Furthermore, in the high density discharges, an extremely high density layer (marfe) exists in front of the antenna before the application of RF, and this layer persists for several milliseconds into the RF pulse. Since the marfe reflects the wave and does not allow its penetration to the bulk plasma, heating is not expected within the framework of standard wave propagation theory. A wave focusing phenomenon is introduced to explain the wave tunnelling through the marfe, using a localized collisional heating combined with the constraint of constant pressure along the magnetic flux line. After the marfe has been removed, after several milliseconds from the RF onset, the power deposition for the bulk plasma appears to be at large radius, according to the soft X-ray diode array diagnostics. But whether the deposition is at about r/a ∼ 2/3 or at the very edge of the plasma has not been confirmed. Theoretically, the accessibility of the overdense plasma for the wave can be improved by introducing a modest level of low frequency density fluctuations into the ray tracing calculations. The damping rate is not, however, large in the inner portion of the plasma. The heating is predicted to take place at large radius, r/a > 0.8, either by heating at the gyroresonance layer after multiple wave reflections on the vessel wall or by a parametric decay instability which excites an ion acoustic mode. Thus, contrary to our intuitive expectation that the deposition might take place in the plasma interior, we are led to conclude that the observed good heating is, indeed, due to the edge heating. The gross energy confinement time τE evaluated at r/a = 2/3 is a factor of about 2.3 larger than the global value τE(a), because of the lack of ECH power deposition within that radius. Since more than 85% of total plasma kinetic energy is contained within r/a = 2/3, this edge heating appears to be very effective in raising the bulk plasma energy.

Predictions of electron cyclotron current drive efficiency for a top-launched extraordinary mode in a Tokamak

A relativistic ray tracing code is used in conjunction with a Fokker-Planck code to calculate the dependence of current drive efficiency on plasma parameters and wave characteristics for top launch. Physical considerations include the interplay of wave polarisation, relativistic cyclotron resonance, optical path length, and electron collisionality. The current drive efficiency at a given temperature is a function of only two parameters: electron number density, and the parameter beta identical to ( omega - Omega 0)/k/sub /// nu B, which includes information on magnetic field strength, wave launch angle and plasma temperature. Peak current drive efficiencies can exceed by 25% the values for comparable inside launch configurations.

Electron cyclotron heating experiments on the JFT-2 tokamak using an inside launch antenna

An electron cyclotron heating experiment is described in which 28 GHz microwave power is launched from the high field side (inside) of a tokamak discharge from a steerable phased array antenna exciting the extraordinary mode. The central temperature was doubled (from 600 eV to 1200 eV) with a power at the antenna approximately two-thirds the Ohmic input. An oblique launch of the extraordinary mode was found to heat more efficiently and to a higher density than either the perpendicularly launched extraordinary mode or the ordinary mode launched from the outside.