G. Tynan

A new plasma-surface interactions research facility: PISCES-B and first materials erosion experiments on bulk-boronized graphite

A new plasma‐surface interactions research facility, PISCES‐B, has been designed and constructed at the University of California, Los Angeles (UCLA). The entire vacuum chamber is bakable and a base pressure of the order of 10−8 Torr is attainable. The PISCES‐B facility can generate continuous plasmas of argon, helium, hydrogen, deuterium, and nitrogen. The density of these plasmas ranges from 1×1011 to 3×1013 cm−3 and the electron temperature ranges from 3 to 51 eV. The plasma bombarding flux to the target can be varied from 1×1017 to 8×1018 ions cm−2u2009s−1. The neutral pressure is controllable in the range from 3×10−5 to 1×10−3 Torr during plasma operation. An in situ surface analysis station with Auger electron spectroscopy (AES), x‐ray photoemission spectroscopy (XPS), and secondary ion mass spectroscopy capabilities is attached to the main plasma experimental chamber. Using the PISCES‐B facility, first materials erosion experiments have been conducted on newly developed bulk‐boronized graphites and sele...

Turbulent edge transport in the Princeton Beta Experiment‐Modified high confinement mode

The first probe measurements of edge turbulence and transport in a neutral beam induced high confinement mode (H‐mode) are reported. A strong negative radial electric field is directly observed in H‐mode. A transient suppression of normalized ion saturation and floating potential fluctuation levels occurs at the low confinement mode to high confinement mode (L–H) transition, followed by a recovery to near low mode (L‐mode) levels. The average poloidal wave number and the poloidal wave‐number spectral width are decreased, and the correlation between fluctuating density and potential is reduced. A large‐amplitude coherent oscillation, localized to the strong radial electric field region, is observed in H‐mode but does not cause transport. In H‐mode the effective turbulent diffusion coefficient is reduced by an order of magnitude inside the last closed flux surface and in the scrape‐off layer. The results are compared with a heuristic model of turbulence suppression by velocity‐shear stabilization.

Experimental simulation of the gaseous divertor concept in PISCES-A

Abstract The concept of the reentrant divertor (or gaseous divertor) has been suggested as a possible solution to the divertor heat load problem in next generation tokamaks. The idea of the reentrant divertor is to redistribute the divertor heat flux over a large surface area by radiation and/or elastic and inelastic collisions with neutral particles. Simulation experiments are performed in the PISCES-A linear plasma device to test the basic concept and to evaluate the axial and radial particle and heat transport. To date, data have been obtained in steady state hydrogen and argon plasmas at densities of up to 2 × 10 13 cm −3 and electron temperatures of 5–30 eV. With moderate gas feed (10 mTorr) to a simulated divertor slot (length 90 cm) we have observed the electron temperature to decrease axially from 25 eV to 3 eV. At higher neutral pressure (> 25 mTorr) a neutralizer regime is found, where the plasma density at the simulated divertor target can be reduced by more than two orders of magnitude. Radial plasma loss is proportional to the neutral pressure and greatly enhanced as compared to the Bohm rate and the classical diffusion rate. The axial plasma heat flux to the divertor target is reduced by a factor of up to 2 × 10 3 .

Electrostatic biasing of the ALT-II pump limiter

Electrostatic biasing experiments using the Advanced Limiter Test (ALT-II) pump limiter in the TEXTOR tokamak have been carried out with the dual goals of: (a) improving the core plasma confinement in the tokamak and (b) enhancing the performance of the pump limiter. The fully toroidal belt limiter has been biased during both ohmic and neutral beam heated discharges. Both polarities of bias have been applied up to a maximum of +or-500 V with no evidence of impurity accumulation in the central plasma, although applying either polarity of bias to the limiter increases recycling from both the limiter face and the vacuum vessel liner. This in turn results in an increase of the central density. The application of a negative bias to the limiter produces a barrier to radial particle transport in the region between the limiter and the wall. This barrier is not observed in either the no bias or the positive bias case. Neither polarity of limiter bias affects the central plasma energy confinement, apparently because the electric field structure producing the radial barrier is outside the limiter tangency radius. The enhanced recycling, coupled with high edge density, increases the radiated power from the plasma edge and may lower the power flux to the plasma facing surface of the limiter blade. In the case of positive limiter biasing, the pressure in the pumped plasma collection scoops of the limiter increases by approximately 20%, corresponding to a similar increase in the particle removal rare of the pump limiter. The increase in the particle removal rate appears to result from a lower edge electron temperature. This is consistent with the observation of an increase in edge radiated power

Presheath profiles in simulated tokamak edge plasmas

Abstract Steady state magnetized plasmas produced by the PISCES experiment are used to study plasma-wall interaction phenomena relevant to confinement devices such as tokamaks. An experimental investigation of the presheath region that extends from a wall surface into “simulated tokamak” edge plasmas along magnetic field lines is reported. Diagnostics especially developed for this work include a fast-scanning multiple Langmuir/Emissive/Mach probe system and a CID camera imaging system. Measurements of density, electron temperature, floating potential, space potential, and bulk plasma flow velocities have been obtained in plasmas with densities ranging from 1012 to 1013 cm−3, electron temperatures from 5 to 15 eV. and axial magnetic fields of 0.2 to 1.4 kG. Plasma density profiles along the magnetic field typically show a characteristic factor of 2 decrease towards the wall surface. A plasma potential variation in the near presheath zone of order 0.5 T e is measured, consistent with the bulk plasma flow approaching the ion sound speed near the wall surface, as inferred from a simple “free fall” model. A Boltzmann model for the presheath density profile accuracy tracks the density profile measured both by the Mach probe and by spectroscopic means. Flow profiles are used as a consistency check on various magnetized Mach probe theories. Results suggest that cross-field transport of parallel momentum through viscosity is relatively unimportant in PISCES plasmas and thus may be unimportant in tokamak boundary layer plasmas. Discharges with non-thermal electrons display axial profiles of space potential and floating potential which indicate a “hotter” electron distribution function near the wall surface, consistent with “colder” electrons being reflected by the presheath potential drop.

Externally-driven H-mode studies in CCT

Test particle radial flux surface excursions are reduced during the H-mode. Particle transport is reduced by a factor of 10 in the H-mode, but energy confinement increases are small. In the H-mode the evolution of poloidally resolved turbulent statistics are not explained by published theory. Turbulent momentum transport leads to a concentration of poloidal momentum within the transport barrier, and compressibility leads to poloidal shock-like phenomena. The electron distribution functions may be modified by this shock, leading to kinetic instabilities. A physics-based understanding of the H-mode must therefore include toroidal effects combined with an adequate treatment of particle orbits, plasma compressibility and associated kinetic effects, and at least a two-species model of turbulent transport. The results suggest that rapid poloidal core-plasma rotation could form core transport barriers without reliance on fluid shear or reversed magnetic shear effects.

E×B transport in the DIII-D boundary plasma

We have measured the electrostatic turbulence and associated particle transport in the DIII-D boundary plasma using a fast reciprocating Langmuir probe array located on the outboard midplane. Both the normalized rms fluctuation levels (density and floating potential) and the fluctuation-driven particle transport are altered by the L-H transition in the SOL. At the separatrix, the density fluctuation level is reduced a factor of 2, consistent with reflectometry results. There is a corresponding decrease in the turbulent particle flux. Deeper in the SOL, the turbulent particle transport in H-mode exceeds the L-mode value. The perpendicular diffusion coefficient D ⊥ and particle confinement time τ p have been estimated, assuming that the transport is purely turbulent and uniform on a flux surface. We find D ⊥ =0.7 D B (L) and 0.04 D B (ELM-free H), and τ p =54 ms (L) and 480 ms (ELM-free H).

Effects of radial electric fields on the turbulence and transport in the TEXTOR edge and SOL plasma

Results are given on the changes in edge transport in the presence of positive fields, below and above the H-mode threshold, induced in TEXTOR by means of a biasing electrode. Langmuir probes are used to measure the time-stationary convective and turbulence driven particle transport. Experiments with normal and reversed toroidal field directions provide a first indication of the spatial variation of the particle fluxes. The convective flux locally dominates the transport inside the limiter radius, but appears to be strongly position dependent, both in magnitude and direction. The turbulence driven particle flux is more spatially uniform. With respect to a reference ohmic discharge, the convective flux strongly increases inside the limiter when positive bias is applied with only small changes occurring at the H-mode transition. Radial fluxes as high as 2×1021 particles/m2s, inwardly or outwardly directed, can be observed at the limiter radius. The turbulent flux inside the limiter during ohmic discharges is outwards, increases with a +250 V bias voltage, but turns inwards in H-mode.

Impurity transport and retention in a gas target divertor: simulation experiments in PISCES-A and modeling results

Impurity retention in the gaseous divertor regime is investigated in the PISCES-A facility at UCLA. We report measurements and 1 1/2D fluid modeling results of impurity transport for typical tokamak divertor plasma parameters (10 18 ≤ n e ≤3×10 19 m −3 , kT e ≤20 eV). The neutral hydrogen density close to the (simulated) divertor target is 10 20 ≤ n 0 ≤3×10 21 m −3 . Gaseous trace impurities (argon, neon) as well as low- Z and high- Z materials sputtering carbon, tungsten) are studied. It is observed that the impurity retention in a gaseous divertor is substantially improved as compared to conventional divertor operating regimes. The modeling results suggest that the retention of neutral and ionized impurities is mainly due to collisions with hydrogen (deuterium) neutrals and ions streaming towards the divertor target a a velocity of 0.25–0.5 c s . A low level of residual impurity transport, observed at high neutral density, is attributed to a plasma flow reversal close to the radial boundary. Sputtering of a tungsten sample by intrinsic impurities has been shown to decrease substantially for target electron temperatures kT e

An electrostatic barrier scrape-off layer for control of core plasma effluxes in tokamaks

It is observed in many Tokamaks that particle and heat fluxes from the core region are poloidally asymmetric, favoring higher cross-field transport on the large major radius edge of the torus. The authors propose a novel technique that may allow one to control this asymmetric flux into the Tokamak boundary plasma. The scheme principally involves the formation of a mobility limited transport layer or electrostatic barrier to inhibit the flow of plasma into the boundary layer at large major radii, forcing plasma to exit instead on the small major radius side of the torus. The implications of such a scheme are potentially important. By forcing plasma to exit on the inside half of the torus where the intrinsic cross-field transport is lower, the overall confinement characteristics of the central plasma may be significantly improved. Furthermore, scrape-off plasma fluxes, subsequent recycling conditions, and their asymmetries at limiter, divertor, and wall structures can be actively controlled.