T. S. Bigelow

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.

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.

A swept two‐frequency microwave reflectometer for edge density profile measurements on TFTR

The edge density profile can play a significant role in determining the plasma confinement and the coupling of the ion cyclotron resonance frequency (ICRF) heating power to the plasma. To experimentally measure the edge density profile in the Tokamak Fusion Test Facility (TFTR), a two‐frequency microwave reflectometer is being built. This reflectometer will operate in a swept two‐frequency configuration between 91 and 118 GHz using the extraordinary mode. The frequency separation between the two microwave signals will be held constant while the signals are swept across the frequency band. By measuring the differential phase delay between these two signals, the density profile can be reconstructed. Two‐frequency profile reflectometry is discussed and results of modeling of this type of reflectometer measurement for TFTR are shown. Finally, the design of the TFTR edge profile reflectometer microwave system is described.

Three dimensional modelling of ICRF launchers for fusion devices

The three dimensional (3-D) nature of antennas for fusion applications in the ion cyclotron range of frequencies (ICRF) requires accurate modelling to design and analyse new antennas. In this article, analysis and design tools for radiofrequency (RF) antennas are successfully benchmarked with experiment, and the 3-D physics of the launched waves is explored. The systematic analysis combines measured density profiles from a reflectometer system, transmission line circuit modelling, detailed 3-D magnetostatics modelling and a new 3-D electromagnetic antenna model including plasma. This analysis gives very good agreement with measured loading data from the Tokamak Fusion Test Reactor (TFTR) Bay-M antenna, thus demonstrating the validity of the analysis for the design of new RF antennas. The 3-D modelling is contrasted with 2-D models, and significant deficiencies are found in the latter. The 2-D models are in error by as much as a factor of 2 in real and reactive loading, even after they are corrected for the most obvious 3-D effects. Three dimensional effects play the most significant role at low parallel wavenumbers, where the launched power spectrum can be quite different from the predictions of 2-D models. Three dimensional effects should not be ignored for many RF designs, especially those intended for fast wave current drive.

Effect of plasma shaping on performance in the National Spherical Torus Experiment

The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving βt∼40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ∼2.8 and triangularity δ∼0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S≡q95Ip∕(aBt), which has been observed at large values of the S∼37[MA∕(m∙T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping ...

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.

Differential‐phase reflectometry for edge profile measurements on Tokamak fusion test reactor

Edge electron density profile measurements, including the scrape‐off layer, have been made during ion cyclotron range of frequency (ICRF) heating with the two‐frequency differential‐phase reflectometer installed on an ICRF antenna on the Tokamak fusion test reactor (TFTR). This system probes the plasma using the extraordinary mode with two signals swept from 90 to 118 GHz, while maintaining a fixed‐difference frequency of 125 MHz. The extraordinary mode is used to obtain density profiles in the range of 1×1011–3×1013 cm−3 in high‐field (4.5–4.9 T) full‐size (R0=2.62 m, a=0.96 m) TFTR plasmas. The reflectometer launcher is located in an ICRF antenna and views the plasma through a small penetration in the center of the Faraday shield. A 26‐m‐long overmoded waveguide run connects the launcher to the reflectometer microwave electronics. Profile measurements made with this reflectometer system will be presented along with a discussion of the characteristics of this differential phase reflectometer and data ana...

Second stability in the ATF torsatron—Experiment and theory

Access to the magnetohydrodynamic (MHD) second stability regime has been achieved in the Advanced Toroidal Facility (ATF) torsatron [Fusion Technol. 10, 179 (1986)]. Operation with a field error that reduced the plasma radius and edge rotational transform resulted in peaked pressure profiles and increased Shafranov shift that lowered the theoretical transition to ideal MHD second stability to β0≊1.3%; the experimental β values (β0≤3%) are well above this transition. The measured magnetic fluctuations decrease with increasing β, and the pressure profile broadens, consistent with the theoretical expectations for self‐stabilization of resistive interchange modes. Initial results from experiments with the field error removed show that the pressure profile is now broader. These later discharges are characterized by a transition to improved (×2–3) confinement and a marked change in the edge density fluctuation spectrum, but the causal relationship of these changes is not yet clear.

Recent results from the ATF torsatron

Recent experiments in the Advanced Toroidal Facility (ATF) torsatron [Plasma Physics and Controlled Nuclear Fusion Research 1990 (IAEA, Vienna, in press)] have emphasized the role of magnetic configuration control in transport studies. Long‐pulse plasma operation up to 20 sec has been achieved with electron cyclotron heating (ECH). With neutral beam injection (NBI) power of ≥1 MW, global energy confinement times of 30 msec have been obtained with line‐average densities up to 1.3×1020 m−3. The energy confinement and the operational space in ATF are roughly the same as those in tokamaks of similar size and field. The empirical scaling observed is similar to gyro‐reduced Bohm scaling with favorable dependences on density and field offsetting an unfavorable power dependence. The toroidal current measured during ECH is identified as the bootstrap current. The observed currents agree well with predictions of neoclassical theory in magnitude and in parametric dependence. Variations of the magnetic configuration ...

Non-inductive solenoid-less plasma current startup in NSTX using transient CHI

Coaxial Helicity Injection (CHI) has been successfully used in the National Spherical Torus Experiment (NSTX) for a demonstration of closed flux current generation without the use of the central solenoid. The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now produced closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.