R.A. Ellis

Maturing ECRF technology for plasma control

The availability of high power (~1 MW), long pulse length (effectively cw), high frequency (>100 GHz) gyrotrons has created the opportunity for enhanced scientific results on magnetic confinement devices for fusion research worldwide. This has led to successful experiments on electron cyclotron heating, electron cyclotron current drive, non-inductive tokamak operation, tokamak energy transport, suppression of instabilities and advanced profile control leading to enhanced performance. The key development in the gyrotron community that has led to the realization of high power long pulse gyrotrons is the availability of edge cooled synthetic diamond gyrotron output windows, which have low loss and excellent thermal and mechanical properties. In addition to the emergence of reliable high power gyrotrons, ancillary equipment for efficient microwave transmission over distances of hundreds of metres, polarization control, diagnostics, and flexible launch geometry have all been developed and proved in regular service.

Fast-ion effects during test blanket module simulation experiments in DIII-D

Fast beam-ion losses were studied in DIII-D in the presence of a scaled mock-up of two test blanket modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot, predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot, which is predicted to be different among the various codes.

Effects of electron trapping and transport on electron cyclotron current drive on DIII-D

Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The experimental ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co- and counter-injection; the ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (TORAY-GA) as well as a quasilinear Fokker–Planck model (CQL3D) and is found to be in better agreement with the more complete Fokker–Planck calculation, especially when the rf power density and/or loop voltage exceed criterion for substantial nonlinear modification of the electron distribution function. The width of the measured ECCD profile is consistent with the theoretically expected width in the absence of radial transport for the current carrying electrons.

Progress toward commissioning and plasma operation in NSTX-U

The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5–10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2–3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3–6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

Launcher performance and thermal capability of the DIII-D ECH system

The temperatures of components of DIII-D ECH launchers were observed during 2003 tokamak operation. The injected power was typically 500-700 kW and the pulse length was typically 2 s. Plasma shots were performed at intervals of about 17 min from 9 a.m. to 5 p.m. The temperatures of a movable mirror, a fixed mirror and a launcher reached an equilibrium after about six hours of repetitive pulsing. The saturation temperature depends to some extent on the plasma stored energy. However, even in high /spl beta/ plasma, the temperatures plateaued at acceptable values.

Making of the NSTX facility

The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.

Physics design of a 28 GHz electron heating system for the National Spherical Torus experiment upgrade

A megawatt-level, 28 GHz electron heating system is being designed to support non-inductive (NI) plasma current (Ip) start-up and local heating and current drive (CD) in H-mode discharges in the National Spherical Torus Experiment Upgrade (NSTX-U). The development of fully NI Ip start-up and ramp-up is an important goal of the NSTXU research program. 28 GHz electron cyclotron (EC) heating is predicted to rapidly increase the central electron temperature (Te(0)) of low density NI plasmas generated by Coaxial Helicity Injection (CHI). The increased Te(0) will significantly reduce the Ip decay rate of CHI plasmas, allowing the coupling of fast wave heating and neutral beam injection. Also 28 GHz electron Bernstein wave (EBW) heating and CD can be used during the Ip flat top in NSTX-U discharges when the plasma is overdense. Ray tracing and Fokker-Planck numerical simulation codes have been used to model EC and EBW heating and CD in NSTX-U. This paper presents a pre-conceptual design for the 28 GHz heating sy...

MODELING RESULTS FOR PROPOSED NSTX 28 GHZ ECH/EBWH SYSTEM

A 28 GHz electron cyclotron heating (ECH) and electron Bernstein wave heating (EBWH) system has been proposed for installation on the National Spherical Torus Experiment (NSTX). A 350 kW gyrotron connected to a fixed horn antenna is proposed for ECH-assisted solenoid-free plasma startup. Modeling predicts strong first pass on-axis EC absorption, even for low electron temperature, Te ~ 20 eV, Coaxial Helicity Injection (CHI) startup plasmas. ECH will heat the CHI plasma to Te ~ 300 eV, providing a suitable target plasma for 30 MHz high-harmonic fast wave heating. A second gyrotron and steered O-X-B mirror launcher is proposed for EBWH experiments. Radiometric measurements of thermal EBW emission detected via B-X-O coupling on NSTX support implementation of the proposed system. 80% B-X-O coupling efficiency was measured in L-mode plasmas and 60% B-X-O coupling efficiency was recently measured in H-mode plasmas conditioned with evaporated lithium. Modeling predicts local on-axis EBW heating and current drive using 28 GHz power in β ~ 20% NSTX plasmas should be possible, with current drive efficiencies ~ 40 kA/MW.

Upgrade to the ECH System on the DIII-D Tokamak

A significant upgrade to the 110 GHz DIII-D ECH system was begun in 2005. Two short pulse (< 2 s) GYCOM gyrotrons were replaced by CPI diode gyrotrons having acceptance parameters of 1.0 MW, 5 s and designed pulse length of 10 s. In addition, a CPI gyrotron with depressed collector, which has demonstrated 1.3 MW for short pulses, was installed and is being tested. Two CPI gyrotrons, which have experienced collector failures, are being repaired.

The 110 GHz Microwave Heating System on the DIII‐D Tokamak

OAK-B135 Six 110 GHz gyrotrons in the 1 MW class are operational on DIII-D. Source power is > 4.0 MW for pulse lengths {le} 2.1 s and {approx} 2.8 MW for 5.0 s. The rf beams can be steered poloidally across the tokamak upper half plane at off-perpendicular injection angles in the toroidal direction up to {+-} 20{sup o}. measured transmission line loss is about -1 dB for the longest line, which is 92 m long with 11 miter bends. Coupling efficiency into the waveguide is {approx} 93% for the Gaussian rf beams. The transmission lines are evacuated and windowless except for the gyrotron output window and include flexible control of the elliptical polarization of the injected rf beam with remote controlled grooved mirrors in two of the miter bends on each line. The injected power can be modulated according to a predetermined program or controlled by the DIII-D plasma control system using real time feedback based on diagnostic signals obtained during the plasma pulse. Three gyrotrons have operated at 1.0 MW output power for 5.0 s. Peak central temperatures of the artificially grown diamond gyrotron output windows are < 180 C at equilibrium.