J. B. O. Caughman

Pellet fuelling and control of burning plasmas in ITER

Pellet injection from the inner wall is planned for use in ITER as the primary core fuelling system since gas fuelling is expected to be highly inefficient in burning plasmas. Tests of the inner wall guide tube have shown that 5 mm pellets with up to 300 m s−1 speeds can survive intact and provide the necessary core fuelling rate. Modelling and extrapolation of the inner wall pellet injection experiments from present days smaller tokamaks leads to the prediction that this method will provide efficient core fuelling beyond the pedestal region. Using pellets for triggering of frequent small edge localized modes is an attractive additional benefit that the pellet injection system can provide. A description of the ITER pellet injection systems capabilities for fuelling and ELM triggering is presented and performance expectations and fusion power control aspects are discussed.

Growth of vertically aligned carbon nanofibers by low-pressure inductively coupled plasma-enhanced chemical vapor deposition

Vertically aligned carbon nanofibers (VACNFs) have been grown using a low-pressure, plasma-enhanced, chemical vapor deposition process. The nanofibers are grown from a nickel catalyst that can be patterned to form arrays of individual, isolated VACNFs. The fibers are grown at pressures below 100 mTorr, using an inductively coupled plasma source with a radio-frequency bias on the sample substrate to allow for independent control of the ion energies. Plasma conditions are related to growth results by comparing optical emission from the plasma to the physical structure of the nanofibers. We find that the ratio of etching species in the plasma to depositing species is critical to the final shape of the carbon structures that are formed.

Loss Estimate for ITER ECH Transmission Line Including Multimode Propagation

Abstract The ITER electron cyclotron heating (ECH) transmission lines (TLs) are 63.5-mm-diam corrugated waveguides that will each carry 1 MW of power at 170 GHz. The TL is defined here as the corrugated waveguide system connecting the gyrotron mirror optics unit (MOU) to the entrance of the ECH launcher and includes miter bends and other corrugated waveguide components. The losses on the ITER TL have been calculated for four possible cases corresponding to having HE11 mode purity at the input of the TL of 100, 97, 90, and 80%. The losses due to coupling, ohmic, and mode conversion loss are evaluated in detail using a numerical code and analytical approaches. Estimates of the calorimetric loss on the line show that the output power is reduced by about 5, ±1% because of ohmic loss in each of the four cases. Estimates of the mode conversion loss show that the fraction of output power in the HE11 mode is ~3% smaller than the fraction of input power in the HE11 mode. High output mode purity therefore can be achieved only with significantly higher input mode purity. Combining both ohmic and mode conversion loss, the efficiency of the TL from the gyrotron MOU to the ECH launcher can be roughly estimated in theory as 92% times the fraction of input power in the HE11 mode.

Pellet fuelling, ELM pacing and disruption mitigation technology development for ITER

Plasma fuelling with pellet injection, pacing of edge localized modes (ELMs) by small frequent pellets and disruption mitigation with gas jets or injected solid material are some of the most important technological capabilities needed for successful operation of ITER. Tools are being developed at the Oak Ridge National Laboratory that can be employed on ITER to provide the necessary core pellet fuelling and the mitigation of ELMs and disruptions. Here we present progress on the development of the technology to provide reliable high throughput inner wall pellet fuelling, pellet ELM pacing with high frequency small pellets and disruption mitigation with gas jets and shattered pellets. Examples of how these tools can be employed on ITER are discussed.

THE EC H&CD TRANSMISSION LINE FOR ITER

Abstract The transmission line (TL) subsystem associated with the ITER electron cyclotron heating and current drive system has reached the conceptual design maturity. At this stage the responsibility of finalizing the design has been transferred from the ITER Organization to the U.S. Domestic Agency. The purpose of the TL is to transmit the microwaves generated by the 170-GHz gyrotrons installed in the radio-frequency building to the launchers located in one equatorial and four upper tokamak ports. Each TL consists of evacuated HE11 waveguides, direct-current breaks, power monitors, mitre bends, polarizers, switches, loads, and pumping sections and will have a typical length that ranges from 100 to 160 m. Overall transmission efficiency could be as high as 92% depending on the specific path between a given gyrotron and launcher. All components are required to be 2-MW compatible, and their layout and organization have been optimized for simplifying the maintenance accessibility and monitoring the primary tritium barrier integrity. Two different TL layouts are at the moment under study, to accommodate the two alternative options for the European sources: four 2-MW units or eight 1-MW units. In this paper the actual design is presented and the technical requirements are discussed.

Efficient coupling of thermal electron Bernstein waves to the ordinary electromagnetic mode on the National Spherical Torus Experiment

Efficient coupling of thermal electron Bernstein waves (EBW) to ordinary-mode (O-mode) electromagnetic radiation has been measured in plasmas heated by energetic neutral beams and high harmonic fast waves in the National Spherical Torus Experiment (NSTX) [M. Ono, S. Kaye, M. Peng et al., Proceedings of the 17th IAEA Fusion Energy Conference, edited by S. Spak (IAEA, Vienna, Austria, 1999), Vol. 3, p. 1135]. The EBW to electromagnetic mode coupling efficiency was measured to be 0.8±0.2, compared to a numerical EBW modeling prediction of 0.65. The observation of efficient EBW coupling to O mode, in relatively good agreement with numerical modeling, is a necessary prerequisite for implementing a proposed high power EBW current drive system on NSTX.

The Development of the Material Plasma Exposure Experiment

The availability of future fusion devices, such as a fusion nuclear science facility or demonstration fusion power station, greatly depends on long operating lifetimes of plasma facing components in their divertors. ORNL is designing the Material Plasma Exposure eXperiment (MPEX), a superconducting magnet, steady-state device to address the plasma material interactions of fusion reactors. MPEX will utilize a new highintensity plasma source concept based on RF technology. This source concept will allow the experiment to cover the entire expected plasma conditions in the divertor of a future fusion reactor. It will be able to study erosion and redeposition for relevant geometries with relevant electric and magnetic fields in-front of the target. MPEX is being designed to allow for the exposure of a priori neutron-irradiated samples. The target exchange chamber has been designed to undock from the linear plasma generator such that it can be transferred to diagnostics stations for more detailed surface analysis. MPEX is being developed in a staged approach with successively increased capabilities. After the initial development step of the helicon source and electron cyclotron heating system, the source concept is being tested in the Proto-MPEX device. Proto-MPEX has achieved electron densities of more than 4×1019 m-3 with a large diameter (13 cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared with the helicon plasma only without microwave heating.

The Development of Plasma-Material Interaction Facilities for the Future of Fusion Technology

Abstract A new era of fusion research has started with ITER being constructed and DEMO for power demonstration on the horizon. However, the fusion nuclear science needs to be developed before DEMO can be designed. One of the most crucial and most complex outstanding science issues to be solved is the plasma surface interaction (PSI) in the hostile environment of a nuclear fusion reactor. Not only are materials exposed to unprecedented steady-state and transient power fluxes, but they are also exposed to unprecedented neutron fluxes. Both the ion fluxes and the neutron fluxes will change the micro-structure of the plasma facing materials significantly even to the extent that their structural integrity is compromised. New devices have to be developed to address the challenges ahead. Linear plasma-material interaction facilities can play a crucial role in advancing the plasma-material interaction science and the development of plasma facing components for future fusion reactors.

Investigation of electron Bernstein wave (EBW) coupling and its critical dependence on EBW collisional loss in high-β, H-mode ST plasmas

High-β spherical tokamak (ST) plasma conditions cut off propagation of electron cyclotron (EC) waves used for heating and current drive in conventional aspect ratio tokamaks. The electron Bernstein wave (EBW) has no density cutoff and is strongly absorbed and emitted at the EC harmonics, allowing EBWs to be used for heating and current drive in STs. However, this application requires efficient EBW coupling in the high-β, H-mode ST plasma regime. EBW emission (EBE) diagnostics and modelling have been employed on the National Spherical Torus Experiment (NSTX) to study oblique EBW to O-mode (B–X–O) coupling and propagation in H-mode plasmas. Efficient EBW coupling was measured before the L–H transition, but rapidly decayed thereafter. EBE simulations show that EBW collisional damping prior to mode conversion (MC) in the plasma scrape off reduces the coupling efficiency during the H-mode phase when the electron temperature is less than 30 eV inside the MC layer. Lithium evaporation during H-mode plasmas was successfully used to reduce this EBW collisional damping by reducing the electron density and increase the electron temperature in the plasma scrape off. Lithium conditioning increased the measured B–X–O coupling efficiency from less than 10% to 60%, consistent with EBE simulations.

Alternative Techniques for Injecting Massive Quantities of Gas for Plasma-Disruption Mitigation

Injection of massive quantities of noble gases or D2 has proven to be effective at mitigating some of the deleterious effects of disruptions in tokamaks. Two alternative methods that might offer some advantages over the present technique for massive gas injection are ¿shattering¿ massive pellets and employing close-coupled rupture disks. Laboratory testing has been carried out to evaluate their feasibility. For the study of massive pellets, a pipe-gun pellet injector cooled with a cryogenic refrigerator was fitted with a relatively large barrel (16.5-mm bore), and D2 and Ne pellets were made and were accelerated to speeds of ~ 600 and 300 m/s, respectively. Based on the successful proof-of-principle testing with the injector and a special double-impact target to shatter pellets, a similar system has been prepared and installed on DIII-D, with preliminary experiments already carried out. To study the applicability of rupture disks for disruption mitigation, a simple test apparatus was assembled in the laboratory. Commercially available rupture disks of 1-in nominal diameter were tested at conditions relevant for the application on tokamaks, including tests with Ar and He gases and rupture pressures of ~ 54 bar. Some technical and practical issues of implementing this technique on a tokamak are discussed.