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. Improvements and regulation of the facilities and developments of the instruments were performed. The construction for the divertor modification from the original open type to the W-shaped semi-closed type for improving the particle control was finished in May 1997. The modification intends to investigate effects of divertor geometry on divertor functions such as particle and impurity controls, and to realize radiative divertor compatible with good confinement.

With respect to the negative-ion based NBI, input power to the plasma was gradually increased along with improvement of operational optimization to attain 5.2 MW at 350 keV with deuterium negative ion beams and 4.2 MW at 360 keV with hydrogen negative ion beams.

Experiments simulating the helium exhaust in ITER were performed with the W-shaped pumped divertor. Helium atoms introduced in the ELMy-H plasmas for 6 sec by helium neutral beam injection were efficiently exhausted by helium pumping with Ar frosted cryopumps in the divertor. The experiments successfully demonstrate high helium exhaust capability of t_He^*/t_E/FONT>4 in steady state, which satisfies the ITER requirement. These results strongly support the divertor design of ITER.

Because long heating time with a total heating energy of 203 MJ was achieved without harmful increase in impurity and particle recycling, a DT equivalent fusion gain of Q_DT^eqサ0.1 was sustained for 9 sec in a ELMy-H mode.

Toward the advanced feedback controls of multiple parameters, the JT-60U started new feedback controls of central line density and divertor neutral gas pressure in addition to the existing controls of off-axis line density, radiation power and neutron production rate. Characteristics of halo current during disruptions were also studied. The NNB-driven current was identified directly from the internal magnetic measurement and driven current profile was confirmed to be consistent with the ACCOME calculation. The current profile control with LHCD successfully sustained the internal transport barrier in reversed shear plasmas. Continuous TAE modes were observed with NNB for the first time in the world as beam-driven TAE modes.

Objectives of the JFT-2M program are (1) advanced and basic researches for the development of high-performance plasmas for nuclear fusion and (2) contribution to the physics R&D for ITER, taking full advantage of flexibility of a medium-size device.

In the closed divertor experiments, it is found that the closer the divertor geometry becomes, the wider the high confinement regime coexistent with a dense and cold divertor plasma results. A compact toroid (CT) injection system has been installed in collaboration with the Himeji Institute of Technology for the development of the advanced fueling for fusion reactors, such as ITER. Encouraging results were obtained with initial CT injection experiments, such that reduction of radiation loss power was observed after the CT injection into OH plasmas. A heavy ion beam probe system, which was developed by the National Institute for Fusion Science has been installed for clarifying mechanism of improved confinement more definitely through fast measurements of the electric field.

The primary objective of theory and analysis is to improve the physical understanding of the magnetically confined tokamak plasma. Remarkable progress has been made on physical understanding of the reduced transport and the stability not only of ideal MHD modes but also of kinetic ballooning mode in reversed shear plasmas. Progress was also made on the neoclassical transport calculation by the Matrix Inversion method and on the scaling law of an offset nonlinear form for the ELMy-H-mode confinement. A five-point model for the scrape-off layer and divertor plasmas was developed and the inside/outside divertor asymmetry was investigated.

The main focus of the NEXT (Numerical Experiment of Tokamak) project is to simulate tokamak plasmas using particle and fluid models on the developing technology of massively parallel computers. A particle-fluid hybrid model was developed for simulation of the kinetic MHD instabilities. The self-generated radial electric field derived by the Reynolds stress and its effect on transport have been studied to contribute to understanding of improvement of the confinement.

R&D of fusion reactor technology has been focused on the ITER/EDA-related area. Major highlights in FY1997 are as follows.

Winding and heat treatment of Nb₃Sn conductors of all eight layers for the outer module of ITER CS model coil have been successfully completed and the assembling technology for the model coil has been developed. Production of 46 kA cable of Nb₃Al strands for the insert coil was also completed.

Fabrication of two full-scale 1/40-sector models of the ITER vacuum vessel was completed in the end of September 1997. The cross section is D-shape of 15 m high and 9 m wide. Both sectors, each of which uses different fabrication procedures and welding techniques, satisfy a dimensional accuracy of within ア3 mm. Hot isostatic pressing (HIP) technology was developed to fabricate a prototype mock-up of ITER shield blanket modules. Regarding the development of ITER divertor, full scale mock-ups of the vertical plates and the wings were successfully fabricated using newly developed bonding technology. The mock-ups were subjected to thermal cycle tests under an ITER steady-state heat load condition of 5 MW/m². The tests prove that all the mock-ups can endure the heat load for a repetition of 10³ cycles without any damages. As to the development of the ITER blanket handling system, performance tests of the full-scale vehicle system was started for demonstration of remote replacement of 4 ton blanket modules.

A stable negative hydrogen ion beam of 25 mA has been successfully accelerated to 1 MeV with a five-staged accelerator. Development of a gyrotoron has progressed to deliver a maximum energy of 520 kW for 5 sec at 170 GHz with a diamond disk window. A large caisson of 12 m³ was installed in a room of TPL to investigate tritium behavior released into confinement system and environment.

In the fusion reactor design, the DREAM design activity was focused on the prototype reactor. In the area of safety research, safety evaluation code development, LOVA and ICE experiments using small scale models, and the study of tokamak dust removal methods were also carried out.

The Final Design Report (FDR) of ITER was issued by the Director in December 1997. After the review by the Technical Advisory Committee (TAC) in January 1998, the FDR was presented to the ITER Council at its 13th Meeting held in February 1998. The FDR is composed of various technical documents on the detailed design of plasma parameters, tokamak components, plant system and the tokamak building. The major results of safety analyses described in the Non-site Specific Safety Report (NSSR) -2 was also included in the FDR. The technical review of the FDR is being conducted by the four Parties. The Japanese Home Team contributes to the design progress in the various fields through the conduction of design tasks in close collaboration with the Joint Central Team (JCT). The JCT member built up to 161 including 46 Japanese members as of December 1997.