Annual Report of Naka Fusion Research Establishment
from April 1, 1998 to March 31, 1999

Naka Fusion Research Establishment

Japan Atomic Energy Research Institute
Naka-machi, Naka-gun, Ibaraki-ken

(Received August 18, 1999)

This report provides an overview of research and development activities at the Naka Fusion Research Establishment, JAERI, during the period from April 1, 1998 to March 31, 1999. The activities in the Naka Fusion Research Establishment are highlighted by high temperature plasma research in JT-60 and JFT-2M as well as DIII-D (US-Japan collaboration), and progress in ITER EDA, including ITER technology R&D.

The objectives of the JT-60 project are to contribute to the ITER physics R&D and to establish the physics basis for a steady state tokamak fusion reactor like SSTR.

The W-shaped divertor was modified to open the outside divertor slot and improve the divertor pumping in the high-triangularity (d) configuration from November through December 1998 when JT-60 operation was halted. After the integrated tests of a 110 GHz ECH/ECCD system newly introduced in January 1999, the Electron Cyclotron (EC) waves were firstly launched into JT-60 in February 1999.

Highlights of the plasma performance in JT-60 are as follows: (1) the production of a high performance reversed shear plasma with the equivalent fusion amplification factor Q_DT^eq of 1.25 in a deuterium discharge and (2) the sustainment of reversed shear plasma with internal transport barriers in a steady-state for 6 s by the current profile control with lower hybrid waves. High b_p H-mode in a high triangularity configuration was sustained for about 4.5 s at Q_DT^eq of ~0.16 with a large non-inductive current drive fraction of 60-70 % of the plasma current. It should be emphasized that a high normalized beta b_N ~ 2.6-2.7 was sustained in a quasi-steady state in a low-q (q₉₅ ~ 3) regime comparable with ITER due to improved edge stability by high triangularity. The power threshold for L-H transition was reduced by approximately 30 % in the W-shaped divertor compared with the open divertor. Ar gas puffing into an ELMy H-mode plasma increased H-factor up to 1.5 at n_e/n^Gr = 0.7. L-H transition with 350 keV negative ion-based NBI (N-NBI) was first observed in JT-60. With N-NBI, a relatively large driven current of 0.6 MA was demonstrated at the current drive efficiency of 0.6 x 10¹⁹ A/W/m².

JFT-2M carries out advanced and basic researches for the development of high-performance plasmas for nuclear fusion and contribution to the physics R&D for ITER, taking full advantage of flexibility of a medium-size device.

A fast change of the electric potential distribution at the L-H transition was measured directly for the first time with a Heavy Ion Beam Probe in collaboration with National Institute for Fusion Science (NIFS). EC current drive was confirmed and it appeared that no difference in the suppression of the tearing mode between the co-direction and the counter direction current drive. The closed divertor maintained low temperature and high density divertor plasma, and high confinement performance. Fuelling by the compact toroid (CT) injection to a central part of the high power NBI-heated plasma (H-mode) was demonstrated for the first time in collaboration with the Himeji Institute of Technology. Optimized design for the ripple reduction by the ferritic board installation outside the vacuum vessel was established.

The primary objective of theoretical and analytical studies is to improve the physical understanding of the magnetically confined tokamak plasma. Remarkable progress was made on physical understanding of transport phenomena such as the internal barrier in JT-60 in connection with radial electric field. Progress was also made on the study of stability such as the neoclassical tearing mode and the double tearing. As for divertor plasma, the importance of thermoelectric instability was shown by using the five point model for the scrape-off layer and divertor plasmas.

The main focus of the NEXT (Numerical EXperiment of Tokamak) project is to research complex physical processes in core plasmas and divertor plasma by using recently advanced computer resources. The role of shear rotation and weak/reversed magnetic shear on these semi-global toroidal modes was studied by employing our toroidal simulation code together with a non-local theory.

R&D of fusion reactor technology has focused on the ITER EDA-related areas. Major highlights in FY1998 are as follows:
(1) The high precision assembly and electric joint connection of eight layers of solenoid coils to produce the Outer Module of the CS Model Coil have been successfully completed. Successful heat treatment of the CS Insert Coil for activation of Nb3Sn was completed in collaboration with U.S.A.
(2) Fabrication of the full-scale vacuum vessel sector was completed as well as the arrival of the port structure from Russia and remote assembly test plan was in progress. As for seismic isolation development the sub-scaled rubber bearings with diameters from 0.2 m to 0.7 m were tested and basic performance test was completed. Regarding the development of ITER divertor, full-scale mock-ups were fabricated. 5 MW/m² steady state and 20 MW/m² transient heat load tests were also conducted successfully with the 1-m long ion source at JAERI. The divertor mock-ups were shown to withstand the lifetime worthy of 3 years. Full-scale mock-up test for blanket remote replacement based on the rail-mounted vehicle type manipulator system was operated at an automatic control mode. The mock-up weighing 4 tons was fixed in position within accuracy of /FONT>2 mm. A small-scale (300 mm^w x 200 mm^d) first wall/blanket model made of a reduced activation ferritic steel F82H for DEMO breeding blanket was successfully fabricated by Hot Isostatic Pressing (HIP).
(3) Negative H-ions with high current density of 15 mA/cm² were extracted from 9 apertures of the ITER concept source, called KAMABOKO source, and converged to produce 200 mA negative ion-beam at 700 keV for 1 s. Development of a gyrotron has progressed to deliver maximum energy of 450 kW for 8 s (3.6 MJ) at 170 GHz with a diamond disk window. Tritium gas behavior in the caisson (called CATS) of 12 m³ was tested under various conditions to develop the codes.