Robert V. Budny

Attainment of high confinement in neutral beam heated divertor discharges in the PDX tokamak

Abstract The PDX divertor configuration has recently been converted from an open to a closed geometry to inhibit the return of neutral gas from the divertor region to the main chamber. Since then, operation in a regime with high energy confinement in neutral beam heated discharges (ASDEX H-mode) has been routine over a wide range of operating conditions. These H-mode discharges are characterized by a sudden drop in divertor density and H α emission and a spontaneous rise in main chamber plasma density during neutral beam injection. The confinement time is found to scale nearly linearly with plasma current, but can be degraded due either to the presence of edge instabilities or heavy gas puffing. Detailed Thomson scattering temperature profiles show high values of T c near the plasma edge (∼ 450 eV) with sharp radial gradients (∼ 400 eV/cm) near the separatrix. Density profiles are broad and also exhibit steep gradients close to the separatrix.

Wall conditioning with impurity pellet injection on TFTR

Solid lithium and boron pellets have been injected into TFTR plasmas to improve plasma performance by coating the graphite inner wall bumper limiter with a small amount of lower Z pellet material, which reduces the influx of carbon from the walls and reduces the edge electron density. This new wall conditioning technique has been applied successfully when continued He conditioning discharges, which are normally used for wall conditioning, no longer significantly reduce the carbon and deuterium influxes. The results show that both Li and B pellets significantly improve wall conditioning and lead to 15–20% improvements in supershot plasma performance when injected ≥1 s prior to neutral beam injection in supershot target plasmas. Neutral beam penetration calculations indicate that the lower edge densities resulting from Li or B pellet wall conditioning lead to improved beam penetration. Sputtering yield calculations confirm that the addition of small amounts of Li on a graphite target can significantly reduce the C sputtering yield.

Mechanisms responsible for topographical changes in PLT stainless-steel and graphite limiters

PLT limiters used during Ohmic and neutral-beam heated discharges were melted and eroded. For the steel limiters, the pattern of large (20 cm2) melted areas was 0.1–1.0-mm ripples in a regular array. The various types of damage can be explained by the power and momentum fluxes of bulk plasma and beam ions and by disruptions. The plasma scrape-off distance inferred from the extent of the damage ranges from 5 to 30 mm and is consistent with a radially outward drift velocity of 2 × 104 mms−1. A new model of the plasma scrape-off is presented which explains pattern and extent of the damage.

Calorimeter probe studies of PDX and PLT

Abstract Using a combined electrostatic-calorimeter probe, a comprehensive survey of energy flux in the edge plasmas of low field ohmic and beam-heated PDX and PLT discharges has been made. Ions are shown to carry the majority of the power to floating probes. The intercepted energy is found to increase nearly linearly with applied heating power. The scrape-off distances, ranging from 0.5 cm for a 4 null diverted discharge to 2.5 cm for inside dee diverted discharge, are independent of heating power. Direct evidence is presented for the prompt loss to the walls and limiters in PDX of energetic beam ions.

Isotopic scaling of confinement in deuterium--tritium plasmas

The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in the Tokamak Fusion Test Reactor [Plasma Phys. Controlled Fusion 26, 11 (1984)]. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. Radial profiles of the deuterium and tritium densities are determined from the D–T fusion neutron emission profile. The inferred scalings with average isotopic mass are quite strong, with τE∝〈A〉0.85±0.20, τEthermal∝〈A〉0.89±0.20, χitot∝〈A〉−2.6±0.5, and De∝〈A〉−1.4±0.2 at fixed Pinj. For fixed local plasma parameters χitot∝〈A〉−1.8±0.4 is obtained. The quoted 2σ uncertainties include contributions from both diagnostic errors and shot irreproducibility, and are conservatively constructed to attribute the entire scatter in the regressed parameters to uncertainties in the exponent on plasma mass.

Alfven frequency modes at the edge of TFTR plasmas

An Alfven frequency mode (AFM) is very often seen in TFTR neutral beam heated plasmas as well as in ohmic plasmas. This quasi-coherent mode has so far only been seen on magnetic fluctuation diagnostics (Mirnov coils). A close correlation between the plasma edge density and the mode activity (frequency and amplitude) has been observed, which indicates that the AFM is an edge localized mode with r/a>0.85. No direct impact of this mode on the plasma global performance or on fast ion loss (e.g., the alpha particles in DT experiments) has been observed. This mode is not the conventional TAE (toroidicity induced Alfven eigenmode). The present TAE theory cannot explain this observation. Other possible explanations are discussed

Gas-fueling studies in the PDX tokamak

Measurements of the gas fueling characteristics of the PDX tokamak have been extended in parameter range. An earlier study presented the gas fueling efficiency for H/sub 2/ and D/sub 2/ for the standard PDX divertor configurations with a large conductance between the divertor and main plasma chambers. This study presents the observed variations in H/sub 2/, D/sub 2/, and He neutral pressure and ionization light emission for divertor configurations with a restricted conductance between the divertor and main plasma chambers. The restricted conductance improved the divertor/main-chamber compression ratio by more than an order of magnitude. For the same plasma density, gas fueling from the divertor chamber was twice as efficient as fueling from the main chamber. At the highest plasma densities that were investigated, anti n/sub e/ approx. = 4 x 10/sup 13/ cm/sup -3/, a decrease in the plasma temperature in the divertor was indicated by a decrease in the ionization light at the divertor throat for D/sub 2/ and He fueled discharges. These observations are consistent with a Monte-Carlo model of neutral gas transport in the divertor.

Particle and heat flux measurements in PDX edge plasmas

Abstract This paper describes the use of novel combined Langmuir-calorimeter probes to measure edge plasma conditions near the midplane in PDX. The probes consisted of up to five Langmuir probes and up to two calorimeters. Single and double probe characteristics yield n e and T c which are compared with results derived from a triple probe analysis. The calorimeters measure heat flux in the electron and ion drift directions. This paper presents time-resolved radial profiles of n e , T e , V F (floating potential),and P (heat flux) during high power neutral beam-heated, single-null discharges and circular scoop limiter discharges. The temporal dependence of these quantities displays the previously observed behavior with respect to gross discharge characteristics; however, an additional dependence on confinement mode has been observed. During the H-mode of energy confinement, a transient depression of n e , T e , and P occur in the scrape-off plasma.

Plasma-materials interactions during RF experiments in tokamaks

Plasma-materials interactions studied in recent ICRF heating and lower hybrid current drive experiments are reviewed. The microscopic processes responsible for impurity generation are discussed. In ICRF experiments, improvements in machine operation and in antenna and feedthrough design have allowed efficient plasma heating at RF powers up to 3 MW. No significant loss of energy from the plasma core due to impurity radiation occurs. Lower hybrid current drive results in the generation and maintenance of hundreds of kiloamperes of plasma current carried by suprathermal electrons. The loss of these electrons and their role in impurity generation are assessed. Methods to avoid this problem are evaluated.

Particle fueling and impurity control in PDX

Abstract Fueling requirements and impurity levels in neutral-beam-heated discharges in the PDX tokamak have been compared for plasmas formed with conventional graphite rail limiters, a particle scoop limiter, and an open or closed poloidal divertor. Gas flows necessary to obtain a given density are highest for diverted discharges and lowest for the scoop limiter. Hydrogen pellet injection provides an efficient alternative fueling technique, and a multiple pellet injector has produced high density discharges for an absorbed neutral beam power of up to 600 kW, above which higher speeds or more massive pellets are required for penetration to the plasma core. Power balance studies indicate that 30–40% of the total input power is radiated while ~15% is absorbed by the limiting surface, except in the open divertor case, where 60% flows to the neutralizer plate. In all operating configurations, Z eff usually rises at the onset of neutral beam injection. Both open divertor pl;asmas and those formed on a well conditioned water-cooled limiter have Z eff ⪅ 2 at the end of neutral injection. A definitive comparison of divertors and limiters for impurity control purposes requires longer beam pulses or higher power levels than available on present machines.