C.E. Bush

Transient electron heat diffusivity obtained from trace impurity injection on TFTR

A new method for obtaining a transient (pulse) electron heat diffusivity (χep) in the radial region 0.38 < r/a < 0.56 in TFTR L mode discharges is presented. Small electron temperature perturbations were caused by single bursts of injected impurities which radiated and cooled the plasma edge. A case of iron injection by laser ablation was found to be more definitive than a supporting helium gas puff case. In this new cold pulse method, the authors concentrate on modelling just the electron temperature perturbations, tracked with electron cyclotron emission diagnostics, and on being able to justify separation of the perturbations in space and time from the cooling source. This χep is obtained for these two cases to be χep = (6.0 m2/s ± 35%) ~ 4χe (power balance), which is consistent with but more definitive than results from other studies that are more susceptible to ambiguities in the source profile

β limit disruptions in the Tokamak Fusion Test Reactor

A disruptive β limit (β=plasma pressure/magnetic pressure) is observed in high‐performance plasmas in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Plasma Phys. Controlled Nuclear Fusion 1, 421 (1987)]. The magnetohydrodynamic character of these disruptions differs substantially from the disruptions in high‐density plasmas (density limit disruptions) on TFTR. The high β disruptions can occur with less than a millisecond warning in the form of a fast growing precursor. The precursor appears to be an n=1 kink strongly coupled through finite β effects and toroidal terms to higher m components. It does not have the ‘‘cold bubble’’ structure found in density limit disruptions. The n=1 kink, in turn, appears to excite a ballooning‐type mode that may contribute to the thermal quench.

Enhanced performance of deuterium--tritium-fueled supershots using extensive lithium conditioning in the Tokamak Fusion Test Reactor

In the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] a substantial improvement in fusion performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive conditioning of the limiter with lithium. This combination has resulted in not only significantly higher global energy confinement times than have previously been obtained in high current supershots, but also in the highest central ratio of thermonuclear fusion output power to input power observed to date.

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 ...

Phenomenology of high density disruptions in the TFTR tokamak

Studies of high density disruptions on TFTR, including a comparison of minor and major disruptions at high density, provide important new information regarding the nature of the disruption mechanism. Further, for the first time, an (m,n)=(1,1) cold bubble precursor to high density disruptions has been experimentally observed in the electron temperature profile. The precursor to major disruptions resembles the vacuum bubble model of disruptions first proposed by B.B. Kadomtsev and O.P. Pogutse (Sov. Phys. JETP 38 (1974) 283)

ICH‐induced plasma rotation on TFTR

ICH‐induced toroidal plasma rotation was initially observed from Doppler shift measurements of the FeXXV Kα line using the TFTR horizontal crystal spectrometer for discharges with ion cyclotron heating of 2 MW RF power and without any neutral beam power in a D(H) plasma. The magnitude of plasma rotation depended on the IC resonance location. The observed rotations were in the counter‐plasma‐current direction and of the order of 5×104 m/s. Simple estimates relate the induced rotation to loss of ions from the plasma core. Later experimental results, as recorded by the charge exchange recombination spectrometer during beam blips at the end of the ICH pulse, with up to 6 MW of RF power in a 4He(H) plasma provided the radial profile of rotation.

Transport simulations of TFTR: theoretically based transport models and current scaling

In order to study the microscopic physics underlying observed L-mode current scaling, 1-1/2-d BALDUR has been used to simulate density and temperature profiles for high and low current, neutral beam heated discharges on TFTR with several semi-empirical, theoretically-based models previously compared for TFTR, including several versions of trapped electron drift wave driven transport. Experiments at TFTR, JET and D3-D show that I{sub p} scaling of {tau}{sub E} does not arise from edge modes as previously thought, and is most likely to arise from nonlocal processes or from the I{sub p}-dependence of local plasma core transport. Consistent with this, it is found that strong current scaling does not arise from any of several edge models of resistive ballooning. Simulations with the profile consistent drift wave model and with a new model for toroidal collisionless trapped electron mode core transport in a multimode formalism, lead to strong current scaling of {tau}{sup E} for the L-mode cases on TFTR. None of the theoretically-based models succeeded in simulating the measured temperature and density profiles for both high and low current experiments.

Performance of ICRF-heated D-T plasmas fueled by neutral beam injection in TFTR

The first experiments to heat neutral‐beam‐fueled D‐T plasmas with ICRF were conducted during the past year on TFTR. These experiments were performed with full‐bore (R∼2.62 m) plasmas in the low recycling, ‘‘supershot’’ regime with confinement times exceeding two times the empirical L‐mode value. Up to 6 MW of RF were coupled into plasmas fueled by up to 24 MW of D‐T beam injection. Some discharges utilized all T beam fueling to maximize the effect of the second harmonic tritium absorption. Extensive Li‐pellet injection was employed in this regime in an attempt to reduce wall recycling and improve plasma confinement. Although significant second harmonic tritium heating was seen in some plasmas, many discharges were characterized by a degradation in performance and reactivity early in the neutral beam pulse as a result of enhanced recycling of impurities and deuterium from the carbon tile limiter. The proximity of the outboard limiter appears to have exacerbated attempts to limit this enhanced influx.

Images of Edge Turbulence in NSTX

The 2-D structure of edge plasma turbulence has been measured in the National Spherical Torus Experiment (NSTX) by viewing the emission of the Da spectral line of deuterium. Images have been made at framing rates of up to 250,000 frames/sec using an ultra-high speed CCD camera developed by Princeton Scientific Instruments. A sequence of images showing the transition between L-mode and H-mode states is shown.

Configuration and Heating Power Dependence of Edge Parameters and H-mode Dynamics in National Spherical Torus Experiment (NSTX)

Edge parameters play a critical role in H-mode (high-confinement mode) access, which is a key component of plasma discharge optimization in present-day toroidal confinement experiments and the design of next-generation devices. Because the edge magnetic topology of a spherical torus (ST) differs from a conventional aspect ratio tokamak, H-modes in STs exhibit important differences compared with tokamaks. The dependence of the NSTX (National Spherical Torus Experiment) edge plasma on heating power, including the L-H transition requirements and the occurrence of edge-localized modes (ELMs), and on divertor configuration is quantified. Comparisons between good L-modes (low-confinement modes) and H-modes show greater differences in the ion channel than the electron channel. The threshold power for the H-mode transition in NSTX is generally above the predictions of a recent ITER (International Thermonuclear Experimental Reactor) scaling. Correlations of transition and ELM phenomena with turbulent fluctuations revealed by Gas Puff Imaging (GPI) and reflectometry are observed. In both single-null and double-null divertor discharges, the density peaks off-axis, sometimes developing prominent ears which can be sustained for many energy confinement times, tau subscript E, in the absence of ELMs. A wide variety of ELM behavior is observed, and ELM characteristics depend on configuration and fueling.