Annual Report of Naka Fusion Research Establishment
From April 1, 2001 to March 31, 2002

Naka Fusion Research Establishment

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

(Received September 25, 2002)

This report provides an overview of research and development activities at Naka Fusion Research Establishment, JAERI, including those performed in collaboration with other research establishments of JAERI, during the period from April 1, 2001 to March 31, 2002. The activities in the Naka Fusion Research Establishment are highlighted by high performance plasma researches in JT-60 and JFT-2M, and completion of ITER Engineering Design Activities (EDA) in July 2001, including technology R&D.

Objectives of the JT-60 project are to contribute the physics R&D for ITER and to establish the physics basis for a steady state tokamak fusion reactor like SSTR. In this fiscal year, most of JT-60 experiments have been devoted to the improvement, sustainment and integration of the high plasma performance.

Highlights of JT-60 experiments are summarized as follows;

1. Using high-power negative-ion-based neutral beam (N-NB) injection (5.7 MW, 402 keV), the highest record for the fusion product (n_i(0) τ_E T_i(0) = 3 x 10²⁰ m^-3 s keV) under the full non-inductive current drive condition was obtained in a high-β_p ELMy H-mode discharge.
   
2. In a reversed shear discharge, the DT-equivalent fusion power gain Q_DT^eq above 0.8 was sustained for 0.55s.
   
3. 3) By the extension of the operational region for the high-triangularity configuration, the quasi-steady beta values maintained for longer than 5τ_E were improved to β_N = 3.05 at a low 95% flux surface (q₉₅ ~ 3.7). Long sustainment (7.4 s, 60 τ_E) of a high β_N (~ 2.7) was also demonstrated.
   
4. High confinement (H_89P ~ 2.7, where H_89P is the energy confinement improvement factor compared with the ITER89-P (L-mode) scaling) was achieved at an electron density of ~ 0.7 times Greenwald density in a reversed shear discharge with q₉₅ ~ 5 using lower hybrid range of frequency (LHRF). By Ar injection, high confinement (HH_y2 > 0.9, where HH_y2 is the energy confinement improvement factor compared with the IPB98(y, 2) scaling), high density (~ 0.8 times Greenwald density) and high radiation loss power (~ 0.8 times heating power) were simultaneously obtained in discharges with the outer separatrix strike point located on the divertor dome top.
   
5. Physics on fusion plasmas, i.e. mechanisms of the current hole, formation of internal transport barriers (ITBs) and transport within it, dynamics of density collapse due to ELMs, the mechanism of in-out asymmetry of divertor particle flux, etc., have been studied under reactor-relevant conditions.

The highlights of technological progress in JT-60 are as follows;

1. Feedback control of a steerable injection mirror to control electron cyclotron range of frequency (ECRF) power deposition was successfully tested using JT-60 discharges. A digital integrating system which has "intelligence" to automatically select unsaturated signals during disruptions has been developed.
   
2. A digital phase controller of thyristor converters, which is applicable to JT-60 superconducting tokamak (JT-60SC), has been developed to replace the aged analog controller for the poloidal magnetic field power supplies of JT-60.
   
3. Surface analyses of the first wall exposed to DD plasmas were initiated with various methods. Tritium deposition profiles on the W-shaped divertor tiles were examined by using tritium imaging plate technique, and compared with the triton deposition calculated with orbit following Monte Carlo code (OFMC) code.
   
4. The injection power of the negative-ion based neutral beam system reached 5.8 MW at 400 keV, and the operation with a pulse length of 10 s at 2.6 MW with one ion source was successfully achieved by improving the beam.
   
5. The electron cyclotron (EC) system for JT-60 achieved the world highest injected power of 2.8 MW for 3.6 s at 110 GHz by improving performances of the gyrotron with built-in RF absorber and realigning the transmission line so as to have high efficiencies of 70-80%.

Objectives of the JT-60SC program are to realize the high-beta steady-state operation of reactor-relevant plasmas and to demonstrate the compatibility of the reduced-activation ferritic steel with the plasma. Physics and engineering design of the JT-60SC made progress on the basis of the objectives.

On JFT-2M, advanced and basic research for the development of high performance tokamak plasma has been promoted, making use of the flexibility of a medium-sized device. In this fiscal year, inside wall of the vacuum vessel was fully covered with ferritic steel plates. A fine structure of the magnetic field inside the vacuum vessel was measured using a three-dimensional magnetic field measurement apparatus. In parallel with this program, advanced and basic study on H-mode plasmas and a compact toroid (CT) injection, etc. has been pursued on JFT-2M.

The principal objective of theoretical and analytical studies is to understand physics of tokamak plasmas. The dynamics of internal transport barrier formation and the relation between the core confinement and the L-mode base were investigated. Progress was also made on the study of MHD instabilities. Surveys on the universality of vertical displacement event (VDE), the effect of polarization current on neoclassical tearing mode (NTM), the feasibility of suppressing NTM by electron cyclotron current drive (ECCD) and divertor characteristics in JT-60SC were carried out. The NEXT (Numerical EXperiment of Tokamak) project has been progressed in order to research complex physical processes both in core and in divertor plasmas by using massively parallel computers. Substantial progresses were made in the studies of turbulence and MHD reconnection and codes were developed to analyze divertor transport in a realistic geometry.

Research and development of fusion reactor technologies have been carried out both to assure the technologies required for the construction of ITER and to accumulate technological data base to assure the design of DEMO, which include the development of the blanket for electric power generation and of reduced activation structural materials and their neutron irradiation facility. Major achievements in the area of fusion reactor technologies in this fiscal year are as follows;

1. Superconducting Magnet: A Nb₃Sn test coil, wound with a diameter of 1.5 m by the Efremov Institute, Russia, successfully generated the target magnetic field of 13 T stably at an operating current of 46 kA at JAERI facility.
   
2. Neutral Beam Injection: A hydrogen negative ion beam with an intense current density of 31 mA/cm² was successfully achieved at a low pressure of 0.1 Pa. The ion beam with a current of 37 mA was accelerated up to a high energy of 0.97 MeV.
   
3. Radio Frequency Heating: An advanced wave launching system called "Remote Steering Launcher" successfully demonstrated a steering capability of +/-10 at 500 kW RF beam power.
   
4. Blanket: By a blanket designed with a pebble-bed of Li₂TiO₃, the local tritium breeding ratio of 1.4 was expected, assuring the overall TBR of more than 1 in a fusion device.
   
5. Structural Materials: Radiation hardening of F82H low activation steel was reduced with tempering temperatures above 700C (<835C).
   
6. In the R&D of International Fusion Material Irradiation Facility (IFMIF), an intense ion source was developed and the performance of a liquid Li target was investigated in detail.
   
7. Tritium Technology: An irradiation of ultraviolet laser of 1 J/cm² successfully removed a carbon co-deposited layer with tritium on the plasma-facing components.
   
8. Fusion Neutronics: A micro fission chamber was developed successfully for a neutron monitor with a good response which was required for the design of the ITER.

In July 2001, nine-year ITER EDA was successfully completed. The first comprehensive design of a fusion experimental reactor based on well-established physics and technology was produced through the activities. The results of the design activities have been completely documented by the hierarchically organized ITER Final Design Report (FDR). The EDA R&D activities with extensive industrial involvement had demonstrated that the main ITER components can function properly.
Following the completion of the EDA, "Co-ordinated Technical Activities (CTA)" were started. CTA means technical activities which are deemed necessary to maintain the integrity of the international project, so as to prepare for ITER Joint Implementation. The central missions of CTA are design adaptation to the specific site(s) conditions, preparation of procurement documents, and assurance of the coherence of the ITER project including design control.

In fusion reactor design activities, a conceptual design of a power reactor with tight aspect ratio was newly proposed for cost reduction. Fuel supply by pellet injection and the erosion rate of the first wall by charge exchange neutrals and alpha particles were studied quantitatively in a fusion power reactor. High heat flux first walls, use of fusion power for fuel production and a reduction of radioactive wastes from the DEMO plant were mainly investigated from importance in socio-economic aspects.