[en] The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes - the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria

[en] Single-fluid transport in the plasma scrape-off layer is modeled for poloidal divertor and mechanically limited discharges. This numerical model is one-dimensional along a field line and time-independent. Conductive and convective transport, as well as impurity and neutral source (sink) terms are included. A simple shooting method technique is used for obtaining solutions. Results are shown for the case of the proposed Alcator DCT tokamak

[en] The stability of dee, inverse-dee and square crosssection plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes - the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria. Experimentally, the square is vertically stable and both dee's unstable to a vertical nonrigid axisymmetric shift. The central magnetic axis displacement grows exponentially with a growth time /sup approx./10³ poloidal Alfven times/sup approx./ plasma L/R time. Proper initial positioning of the plasma on the midplane allows passive feedback to nonlinearly restore vertical motion to a small stable oscillation about the center. Experimental poloidal flux plots are produced directly from internal magnetic probe measurements. The PEST code, ignoring passive feedback, predicts all equilibria to be vertically unstable with the square having the slowest growth. With passive feedback, all are stable. Thus experiment and code agree that the square is the most stable shape, but experiment indicates that passive feedback is partially defeated by finite plasma resistivity. In both code and experiment square-like equilibria exhibit a relatively harmless horizontal instability

[en] The most recent meeting of the Scrape-Off Layer (SOL) and Divertor Physics Group of the International Tokamak Physics Activity (ITPA) was held in Lausanne, Switzerland, on October 21-23, 2002 at the CRPP/EFL laboratory. The meeting was hosted by Dr. R. Pitts of that laboratory, with the help of other laboratory staff. There were 23 participants. The meeting format was slightly changed from that of the previous meeting (February 25-27, 2002) in so far that more time was allowed for discussions, thus making the meeting more productive. There were 28 detailed presentations over two and a half days falling into three primary areas: the physics of ELMs and their effects; radial transport and wall recycling; materials issues (T codeposition, use of W for tiles, etc.)

[en] A visual inspection of the full poloidal ring limiters used on Alcator C identifies two directionally asymmetric damage regions: at the inside-top location on the electron side (as defined by plasma current, I/sub p/) and at the inside-bottom location on the ion side. We report direct measurements of both ion and electron edge parameters using Janus, a two-sided, multi-diagnostic edge probe, which also indicates strong directional asymmetries. Higher ion and electron temperatures and densities at the probe location occur on the electron side when the toroidal field (B/sub t/) is antiparallel to I/sub p/. The degree of asymmetry cannot be explained by limiter configuration alone. The direction of B/sub t/ with respect to I/sub p/, and variation of the plasma in-out positions, change the magnitude of asymmetry. A number of mechanisms which may cause the directional asymmetries are examined, including parallel plasma flow and asymmetric perpendicular transport into the collecting flux tubes. Measurements made by DENSEPACK, a full poloidal array of Langmuir probes, show strong poloidal asymmetries in both plasma density and electron temperature throughout the scrape-off region of Alcator C. Such asymmetries may arise from asymmetric perpendicular transport and/or act to drive plasma flows along field lines. Thus, the dependence of the directional asymmetries on the directions of B/sub t/ and I/sub p/ may be explained

[en] The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes -- the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria. Experimentally, the square is vertically stable and both dee's unstable to a vertical nonrigid axisymmetric shift. The central magnetic axis displacement grows exponentially with a growth time approximately 10³ poloidal Alfven times plasma time. Proper initial positioning of the plasma on the midplane allows passive feedback to nonlinearly restore vertical motion to a small stable oscillation. Experimental poloidal flux plots are produced directly from internal magnetic probe measurements

[en] This paper provides a background for the use of Langmuir and gridded energy analyzer probes in diagnosing plasmas with varied characteristics. Theory is illustrated which governs the analysis of data from, and the design of these probes. Several probe analysis techniques and some of their typical problems are presented

[en] This paper presents a characterization of the Alcator C Scrape-Off Layer (SOL) plasma during ICRF hydrogen-minority fast wave heating experiments. The SOL plasma parameters were measured using a multifunctional probe, JANUS, which is capable of simultaneously measuring the ion and electron parameters both parallel and antiparallel with respect to the toroidal magnetic field. The probe data indicate, at low value of injected rf power, there is direct edge heating and density increases at radii greater than that of the antenna Faraday shield. Increasing the injected rf power spreads both the temperature and density increases throughout the edge region, flattening the radial profiles. Varying the position of the resonance layer in the main plasma does not significantly change the effect of ICRF on the SOL parameters. Given this single spatial point characterization of the SOL, a crude estimate of power flow into and through the edge plasma indicate that /approximately/20% of the ICRF power launched from the antenna is absorbed /und directly/ in the SOL plasma. Additional observation of the impurity source rates confirms the conclusions of an earlier paper, which attributed increasing central densities of high-Z impurities to the increase in physical sputtering rate at both the ICRF antenna's Faraday shield and the limiter surface

[en] One of the high heat flux solutions envisioned for ITER is the gas target divertor. This scheme requires high neutral pressure to be sustained in the divertor chamber with a minimal effect on the pressure in the main tokamak chamber (i.e. high gas compression). The neutral gas compression has been studied in the Alcator C-Mod closed divertor under various central and edge plasma conditions. The neutral pressure measured by a fast, in-situ, ionization gauge, installed behind the divertor target plate was compared with the midplane pressure, measured by a shielded Bayard-Alpert gauge. Divertor pressures up to 30 mTorr with compression factors p_div/p_mid ≤ 70 have been observed. It has been found that the neutral pressure in the divertor does not depend strongly on the fueling location but rather on the core plasma density and the resulting divertor plasma regime. Divertor detachment leads to a considerable drop in the compression ratio, suggesting a partial open-quotes unpluggingclose quotes of the divertor volume. An examination of the local particle flux balance in the divertor indicates that the single most important factor determining divertor pressure and compression is the private-flux plasma channel opacity to neutrals

No abstract available