Annual Report of Naka Fusion Research Establishment
from April 1, 2004 to March 31, 2005

Naka Fusion Research Establishment

Japan Atomic Energy Research Institute
Naka-shi, Ibaraki-ken

(Received August 12, 2005)

  This annual report provides an overview of research and development (R&D) activities at Naka Fusion Research Establishment during the period from 1 April, 2004 to 31 March, 2005, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities.

  In the JT-60 research program, the pulse length of the tokamak discharge was extended successfully up to 65 s in FY 2003. In FY 2004, following the successful results, optimization of long pulse discharges was continued in order to explore the boundaries of facility capabilities for the long pulse operation. The pulse length of the negative-ion based neutral beam injection system has reached up to 25 s with an injection power of 1MW. In electron cyclotron wave system, the pulse length has also extended up to 45 s with an RF power of 0.35 MW by using four gyrotrons in a series operation. Sustainment of higher normalized β of βN >2.3 for 22.3 s, or βN >2.5 for 15.5 s has been achieved by exploiting available plasma heating systems. This discharge exhibits not only the high βN, but also high confinement improvement with the H factor of H89P=1.9-2.3 and high normalized fusion performance of GH89PβN/q952 = 0.4-0.5 during the sustainment, where q95 is a safety factor at the edge. G~0.4 corresponds to the fusion energy gain of Q=10 for the ITER standard scenario. The H-mode plasma with H89p~1.4 has been maintained for about 30 s, although degradation of the performance was observed at the later half of the discharge. In the reversed shear plasmas, the operation regime was successfully extended to the density higher than Greenwald density limit, while maintaining high confinement and high radiation loss fraction by tailoring the internal transport barriers of the density and temperature. Demonstration of neoclassical tearing mode stabilization and improvement of plasma performance in the high beta region (βN~3) has been performed using local current drive by the second harmonic electron cyclotron waves. In addition, a real-time control system of safety factor profile has been developed. This system enables spatial control of driven current by adjusting the parallel refractive index of lower-hybrid waves through the change of phase difference between multi-junction launcher modules.

  The design of National Centralized Tokamak (NCT), which is the superconducting modification of JT-60, progressed both in physics and engineering. Machine has been designed to have a wide-range capability of operation in aspect ratio and plasma shape. Engineering design of the main components of superconducting toroidal and poloidal magnetic field coils, vacuum vessel, in-vessel components, and cryostat has been performed to investigate the structure optimization from viewpoints of manufacturing processes, operation and maintenance feasibility.

  A series of the experimental programs on the JFT-2M were completed in FY 2003. In FY 2004, experimental data on the Advanced Material Tokamak Experiment (AMTEX) using the reduced activation ferritic steel (F82H), high performance experiment, characteristics of SOL and divertor plasma and compact toroid injection for fueling have been analyzed and evaluated. Concerning the AMTEX, analysis of high-β experiments with the Ferritic Inside Wall (FIW) facing close to the plasma have shown a wall stabilization effect. By using an MHD equilibrium code, it has been confirmed that the plasma with βN~3.5 is compatible with FIW.

  In the theoretical and analytical researches, significant progress was made in the studies of transport simulation of current hole plasma, role of low order rational q-values in the ITB events, the theory of Alfven eigenmodes in tokamaks, current spike behavior of disruptive plasma, and stability of external MHD modes. In the project of numerical experiment of tokamak (NEXT), the studies of the structure formations in toroidal electron temperature gradient driven turbulence, control of the zonal flow, and formation of current hole also progressed.

  R&Ds of fusion reactor technologies have been carried out both to further improve technologies necessary for ITER construction, and to accumulate technological database to assure the design of fusion DEMO plants. For the design optimization of ITER superconducting magnets, degradation of critical current performances of the Nb3Sn conductors has been experimentally and numerically examined and a new simulation model has been developed to predict degradation behavior in a large current superconductor. For ITER Neutral Beam Injector, MeV-range accelerator R&D is being in progress and the current density has been extended to 100 A/m2. For the further pulse extension and power increase of 170 GHz gyrotron, a built-in radiator at the mode converter has been optimized to improve the efficiency of gyrotron output power and to reduce stray radiation, and pre-program controls of a cathode heater power has been employed to stabilize the beam current and the output power. In the R&Ds on Plasma Facing Components, a screw tube has been developed as a possible option for the ITER divertor. For the design of ITER Test Blanket Module (TBM), two candidates, namely Water Cooled Solid Breeder TBM and Helium Cooled Solid Breeder TBM have been proposed, and elementary technology R&Ds have been progressed for fabrication of the TBM, thermo-mechanical properties of the packed bed, and irradiation technologies. An outline design of an electrochemical hydrogen pump has been carried out as a candidate of the advanced Blanket Tritium Recovery system. Using DT neutrons, neutronics integral experiments have been performed with a blanket mockup at FNS facility to predict the tritium breeding ratio with an error less than 5%. As one of the most promising structural materials for the ITER TBM and DEMO blankets, F82H has been investigated with its neutron irradiation effects using HFIR, JMTR, and so on. In the IFMIF program, transitional activities have been continued.

  In the ITER Program, along the work plan approved on June 2004 under the framework of the ITER Transitional Arrangements, the Design and R&D Tasks have been carried out by the Participant Teams. In FY 2004 JAERI has been in charge of fifty-five Design Tasks that make the implementation of preparing the procurement documents for facilities and equipments that are scheduled to be ordered at an early stage of ITER construction. The site issues have been continuously discussed among the delegations of six parties/area and through the bilateral negotiation between Japan and the EU based on a viewpoint of "a broader approach" concept.

  Finally, in the fusion reactor design studies, the conceptual design of the fusion DEMO plant which is placed beyond ITER has progressed. Three options with different capabilities of center solenoid (CS) coil are studied. Researches on the physics related to the ramp up CS-less reactor, and waste management have progressed.

Editors : Yamamoto, T., Sato, M., Kudo, Y., Shu, W., Yoshida, H.
Keywords ; JAERI, Fusion Research, JT-60, JFT-2M, Fusion Technology, ITER, Fusion Power Demonstration Plants, Fusion Reactor