H-mode investigations in DIII-D

Conference · Fri May 01 04:00:00 EDT 1987

OSTI ID:6283525

Shimada, M; Burrell, K H; Ejima, S; Schissel, D P; Allen, S; Brooks, N H; Bramson, G; Callis, R W; Carlstrom, T N; Colleraine, A P

DIII-D is a large tokamak (major radius R = 1.67 m, minor radius a = 0.67 m), capable of producing vertically elongated limiter and divertor plasmas with high plasma current at modest toroidal field (B/sub T/ = 2.2 T on axis). To date, maximum plasma currents achieved are 3.0 MA in limiter plasmas (elongation in excess of 2.0), 2.0 MA in divertor plasmas (elongation 1.9). The DIII-D divertor is a reactor-compatible open divertor design, with no separate divertor chamber and no coils in vacuum. The divertor plates are two toroidal bands of carbon tiles on the bottom of the vacuum vessel. Deuterium plasmas are heated by hydrogen neutral beams (75 keV, coinjection only) and electron cyclotron heating /1/. Power levels to date are 6 MW (beam) and 1.0 MW (ECH). H-mode discharges, H-mode thresholds, and Edge Localized Modes have been investigated for this system.

• Research Organization: GA Technologies, Inc., San Diego, CA (USA); Japan Atomic Energy Research Inst., Tokai, Ibaraki; Lawrence Livermore National Lab., CA (USA); Johns Hopkins Univ., Baltimore, MD (USA); Hitachi Ltd., Ibaraki (Japan); Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)
• DOE Contract Number: AC03-84ER51044
• OSTI ID: 6283525
• Report Number(s): GA-A-18864; CONF-870629-9; ON: DE87011833
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
700101* -- Fusion Energy-- Plasma Research-- Confinement
Heating
& Production
700102 -- Fusion Energy-- Plasma Research-- Diagnostics
BEAM INJECTION HEATING
CONFINEMENT
DATA
DIVERTORS
ECR HEATING
EXPERIMENTAL DATA
HEATING
HIGH-FREQUENCY HEATING
INFORMATION
LIMITERS
NUMERICAL DATA
PLASMA
PLASMA CONFINEMENT
PLASMA DIAGNOSTICS
PLASMA HEATING
THERMONUCLEAR REACTORS
TOKAMAK TYPE REACTORS