Section 1. Invited paperJT-60 experiments
Abstract
In the last year, ohmic heating experiments were performed in JT-60 for about three months; May–June 1985 and March 1986. The major experimental effort has been concentrated on the divertor configuration in anticipation of H-mode operation of JT-60 in high power supplementary heating experiments starting in August 1986. Stable discharges with a divertor have been obtained at Ip = 1.85 MA and . Radiated power from the main plasma in divertor discharges is reduced to one third of that in material limiter discharges. The electron density has a broad radial profile and the effective safety reaches qeff = 2.5. The neutral gas pressure in the divertor chamber becomes significantly high and the control of particle recycling has been attempted.
References (5)
- JT-60 Team
Plasma Phys. Contr. Fusion
(1986) - JT-60 Team
Cited by (10)
The toroidal belt pump limiter, ALT-II, in the TEXTOR tokamak
1990, Fusion Engineering and DesignA toroidal belt pump limiter has been designed, constructed, installed and operated in the TEXTOR tokamak. The Advanced Limiter TEST-II, or ALT-II, is carried out as a partnership under a joint agreement between the European Community, Japan and the United States. ALT-II is designed to withstand the high heat fluxes (greater than 500 W/cm2) intrinsic to the boundary region of tokamak plasmas while simultaneously removing 5% to 10% of the core plasma efflux. The belt consists of eight segments, each 28 cm wide, 150 cm long, and 1.2 cm thick, and each is covered with shaped graphite tiles. Mounted behind each blade is a “scoop” system which captures plasma flowing past the leading edges and behind the blades. This plasma neutralizes on a collector plate and the resulting neutral gas is pumped. During ohmic discharges, 50 to 60 A of current is collected in the scoops and removal rates up to 0.25 Torr-l/s are achieved. The blades are instrumented with Langmuir probes and thermo couples, and other diagnostics include infrared cameras, Hα-monitors, pressure gauges, helium charge exchange spectroscopy, and helium gas diagnostics. The plasma edge and the operation of the pump limiter system are described.
Operation experiences of the JT-60 first walls during high-power additional heating experiments
1989, Fusion Engineering and DesignJT-60 started its operation in May 1985 with TiC-coated molybdenum or Inconel 625 first walls. They provided very clean surfaces as well as superior plasma characteristics during Joule heating discharges. Though 20 μm-thick TiC coatings showed good adhesion characteristics, melting of the TiC coating and also the molybdenum or Inconel 625 substrate was observed at some specific spots, and an influx of heavy metals to the main plasma was inevitable during discharges.
Initial results of the additional heating experiments showed degrading effects of locally melted TiC-coated molybdenum or Inconel 625 on plasma operation. Therefore, about a half of the TiC-coated first walls were removed and new graphite first walls were installed during the venting period from April to May 1987. The start-up of the discharge conditioning after installation of a significant number of graphite tiles was very rapid. Flexibility in plasma operation was increased, and JT-60 extended the operation region beyond its original specifications.
The graphite first walls of the main chamber performed admirably and maintained their integrity under the conditions of plasma current and additional heating power up to 3.2 MA and 30 MW, respectively. On the other hand, the number of damaged divertor plates was much larger than that expected. The reason of unexpected failure is now under examination.
Present knowledge about the materials behavior in JT-60
1988, Journal of Nuclear MaterialsJT-60 first walls were originally composed of a number of tiles made of either molybdenum or Inconel 625 with 20 μm thick TiC coating. These first walls, operated after the bakeout at a temperature of 350°C, provided very clean surfaces as well as superior plasma characteristics during Joule heating discharges. Damage of the TiC-coated first walls was modest and localized to specific spots. However, initial results of the additional heating experiments showed degrading effects of locally melted TiC-coated molybdenum and Inconel 625 on the plasma operation.
During the venting period from April to May '87 about half of the TiC-coated molybdenum and Inconel 625 was replaced by graphite first walls to prepare for high-power and long-pulse additional heating experiments. After a three-day 300 °c bakeout, 30 h glow discharge cleaning, and 16 h pulsed discharge cleaning at a vessel temperature of around 250°C, a very rapid start-up of the discharge conditioning with plasma current up to 1.0 MA was achieved with 14 shots. Flexibility in plasma operation was increased and the plasma current was increased to 2.9 MA.
Material behavior and materials problems in TFTR
1988, Journal of Nuclear MaterialsThis paper reviews the experience with first-wall materials in TFTR over a 20 month period of operatiori during 1985–1987. Experience with the axisymmetric inner wall limiter, constructed of graphite tiles, is described, including the necessary conditioning procedures needed for impurity and particle control of high power (≤ 20 MW) neutral injection experiments. The thermal effects in disruptions have been quantified and no significant damage to the bumper limiter has occurred as a result of disruptions. Carbon and metal impurity redeposition effects have been quantified through surface analysis of wall samples. Estimates of the tritium retention in the graphite limiter tiles and redeposited carbon films have been made, based on analysis of deuterium retention in removed graphite tiles and wall samples. New limiter structures have been designed using a 2D carbon/carbon () composite material for RF antenna protection. Laboratory tests of the important thermal, mechanical, and vacuum properties of materials are described. Finally, the last series of experiments in TFTR with in-situ surface pumps are discussed. Problems with embrittlement have led to the removal of the getter material from the in-torus environment.
Plasma induced surface modifications at the first wall components of high temperature plasma experiments
1988, Journal of Nuclear MaterialsThe modifications of the vessel wall and limiter surfaces in today's large plasma machines are reviewed with special emphasis on the investigations at JET concerning material erosion, transport and redeposition. This is correlated with the problems of impurity release and hydrogen recycling. The specific roles of the limiters and the vessel walls are discussed and some consequences on the use and the distribution of different materials for in-vessel components in tokamaks are indicated.
Research activities on tokamaks in Japan: JT-60U, JFT-2M, and TRIAM-1M
2002, Fusion Science and Technology