November 1997


After the divertor modification, the systematic experiments for the halo current analyses have been conducted at the plasma current (Ip) <2MA as an ITER physics R&D. The summary of the experiments shows that the ratio of the maximum local halo current to the plasma current (Ihmax/Ip0) is less than 0.52, which satisfies the design condition of ITER (Ihmax/Ip0<0.58).

A careful analysis of transport coefficients was carried out for the reversed shear discharge with the ELMing H-mode edge where the internal transport barrier (ITB) located just inside the minimum safety factor position was sustained for 1.5-2 sec. The analysis shows that the thermal diffusivities both for electron and ion drop quickly across ITB in the quasi steady state.

These results were presented both in the 38th Annual Meeting of American Physics Society Division of Plasma Physics (APS-DPP) and in the 14th anniversary meeting of Japan Society of Plasma Science and Nuclear Fusion Research (Osaka).


Progress on NNB ( Negative ion-based Neutral Beam) injection experiments in JT-60U was presented in the 38th Annual Meeting of APS-DPP, Pittsburgh.

Regarding current drive study, the profile of beam driven current by NNB injection (360 keV, 2-4 MW) was well identified with MSE measurements and EFIT code introduced from DIII-D. The identified NNB-driven current profile agrees reasonably with the predictions by ACCOME code. Controllability of plasma current was confirmed by the measured profiles of beam driven current with both NNB and PNB (Positive ion-based NB: 80 keV, 1.2 MW).

The TAE-like mode activities were observed with the NNB injection (350 keV, ~3 MW) for the first time. The mode frequency was 40-130 kHz and the toroidal mode number was n=1-3. These TAE-like modes show bursting activities similar to the TAE modes observed using tangential neutral beams in TFTR and DIII-D. The volume-averaged beta of NNB-injected ions was ~0.2 %, which was roughly a half of the threshold value of the TAE mode excitation in TFTR and DIII-D.


The following results were presented in the 14th anniversary meeting of Japan Society of Plasma Science and Nuclear Fusion Research and in the APS-DPP meeting in Pittsburgh.
1) Large gas puff is possible in ELMy H-mode with keeping electron density almost constant, which will be advantageous to produce SOL flow.
2) Helium exhaust from the private flux region like ITER was successfully demonstrated in W-shaped divertor.
3) Onset density of X-point MARFE after the divertor modification was decreased to about 60% of Greenwald density limit , while the onset density of X-point MARFE in the open divertor was about 70% of Greenwald density limit.
4) High power and long pulse discharges with input energy up to 203 MJ became possible in the W-shaped divertor.
5) Divertor density in ohmic plus short pulse ICRF heating could be measured with mm-wave interferometer. Sharp drop in divertor electron density was observed at the timing of L-H transition.
6) Hydrocarbon generation estimated by CD band intensity decreased and carbon concentration in the main plasma decreased when divertor pump was active, suggesting that SOL flow might be produced by gas and pump.