R.W. Conn

Plasma surface interaction experimental facility (PISCES) for materials and edge physics studies

PISCES is a laboratory research facility at UCLA for the study of plasma material interactions in continuous operation at particle and heat fluxes expected in advanced fusion experiments and in reactors. The PISCES plasma has parameters similar to the edge plasma in tokamaks and the halo plasma in tandem mirrors. The 10 cm diameter plasma is generated by a reflex arc discharge from a large area La-Mo disc cathode. The plasma from this H, D, or He discharge transports along the 2–4 kG magnetic field to produce a high density (1013 cm−3), low energy (< 10 eV) plasma. Alternatively, the ions can be extracted and accelerated along the magnetic field by a very low energy acceleration-deceleration system to produce a plasma with 50–500 eV ion energy and a density of about 1012 cm−3. Differential pumping in the baffled chamber and plasma pumping during operation maintain the neutral pressure at about 10 −4 Torr in the target region. Initial operation has produced continuous cold hydrogen plasmas for hours at a time with densities in excess of 1013 cm−3. The three major foci of PISCES experimental work include: materials and coatings behavior under continuous plasma bombardment to high fluence; plasma and neutral gas effects important in the vicinity of the plasma interface with in-vessel components and the gas ducting and pumping systems; and the development, testing, and calibration of gas ducting and pumping systems; and the development, testing, and calibration of diagnostics to be used in long pulse plasma material interaction programs on confinement experiments. The continuous, controllable operation and moderately high heat fluxes (50–125 W/cm2) possible in PISCES permit high fluence investigations on edge plasma physics and materials interactions. Initial plasma operation, experimental investigations, and the capabilities of PISCES will be discussed.

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−2 s−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...

Bulk-boronized graphites for plasma-facing components in ITER

Abstract Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt% to 30 wt% have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600 °C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2–3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt% bulk-boronization at temperatures above 1000 °C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt% bulk-boronization of graphite hinders air oxidation nearly completely at 800° C and reduces the steam oxidation rate by a factor of 2–3 at around 1100 and 1350 °C.

The ARIES-I Tokamak Reactor Study †

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

Plasma flow measurements along the presheath of a magnetized plasma

Plasma flow measurements in the presheath have been performed using two types of directional electric ‘‘Mach’’ probes, in the PISCES facility at UCLA [J. Nucl. Mater. 121, 277 (1984)]. A fast scanning versatile probe combination has been developed, which operates simultaneously as a ‘‘magnetized’’ Mach probe, an ‘‘unmagnetized’’ Mach probe (with characteristic probe size greater than and smaller than ion gyroradius, respectively), and an emissive probe. Presheaths have been investigated by inserting a small object at the center of the plasma column. Variations in plasma flow velocity, density, and potential along the presheath have been deduced by fluid and kinetic theories. A comparison is made between Mach numbers obtained from the magnetized probe and the unmagnetized probe. Incorporation of shear viscosity of order ∼0.5nmiD⊥ in the cross‐field transport along the presheath seems best to model the results. The cross‐field diffusivity (D⊥) is found to scale approximately proportional to B−1/2, with magn...

Plasma and neutral dynamics in a simulated tokamak gas target divertor

A stationary, detached ionization front is observed in an experimentally simulated divertor plasma (n≤3×1019 m−3, kTe≤20 eV) interacting with a hydrogen gas target. With a neutral hydrogen density, n0≊2×1021 m−3, the electron temperature at the simulated divertor target is reduced to kTe target≊2.5 eV. Up to 97% of the electron heat flux (≤7 MW/m2) is dissipated by dissociation and ionization losses and hydrogen line radiation. The plasma pressure is observed to peak near the ionization front, and a plasma flow reversal is observed in the region of reversed pressure gradient. Classical momentum flow parallel to the magnetic field and anomalous cross‐field particle transport are found. The plasma flow is strongly damped by ion–neutral collisions and is subsonic. Numerical results from a one‐and‐one‐half dimensional (11/2‐D) coupled plasma–neutral fluid model (incorporating radial particle transport, recycling, and neutral gas injection) agree well with the experimental data, and indicate that the electron ...

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.

Generalized fluid equations for parallel transport in collisional to weakly collisional plasmas

A new set of two‐fluid equations that are valid from collisional to weakly collisional limits is derived. Starting from gyrokinetic equations in flux coordinates with no zero‐order drifts, a set of moment equations describing plasma transport along the field lines of a space‐ and time‐dependent magnetic field is derived. No restriction on the anisotropy of the ion distribution function is imposed. In the highly collisional limit, these equations reduce to those of Braginskii, while in the weakly collisional limit they are similar to the double adiabatic or Chew, Goldberger, and Low (CGL) equations [Proc. R. Soc. London, Ser. A 236, 112 (1956)]. The new set of equations also exhibits a physical singularity at the sound speed. This singularity is used to derive and compute the sound speed. Numerical examples comparing these equations with conventional transport equations show that in the limit where the ratio of the mean free path λ to the scale length of the magnetic field gradient LB approaches zero, ther...

Tokamak pump limiters

Abstract Experiments with pump limiters on several operating tokamaks have established them as efficient collectors of particles. The gas pressure rise within the chamber behind the limiters has been as high as 50 mTorr when there is no internal chamber pumping. Observations of the plasma power distribution over the front face of these limiter modules yield estimates for the scale length of radial power decay consistent with predictions of relatively simple theory. Interaction of the in-flowing plasma with recycling neutral gas near the limiter deflector plate is predicted to become important when the effective ionization mean free path is comparable to or less than the neutral atom mean path length within the throat structure of the limiter. Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this “Z-mode” of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described.

Preliminary design analysis of the ALT-II limiter for TEXTOR

Abstract Installation of a large toroidal belt pump limiter, Advanced Limiter Test II (ALT-II), on the TEXTOR tokamak at Julich, FRG is anticipated for early 1986. This paper discusses the preliminary mechanical design and materials considerations undertaken as part of the feasibility study phase for ALT-II. Since the actively cooled limiter blade is the component in direct contact with the plasma edge, and thus subject to the severe plasma environment, most preliminary design efforts have concentrated on analysis of the blade. The screening process which led to the recommended preliminary design consisting of a dispersion strengthened copper or OFHC copper cover plate over an austenitic stainless steel base plate is discussed. A 1 to 3 mm thick low atomic number coating consisting of a graded plasma-sprayed Silicon Carbide-Aluminum composite is recommended subject to further experiment and evaluation. Thermal-hydraulic and stress analyses of the limiter blade are also discussed.