Annual Report of Naka Fusion Research Establishment
from April 1, 2003 to March 31, 2004

Naka Fusion Research Establishment

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

(Received October 1, 2004)

  This annual report provides an overview of research and development (R&D) activities at Naka Fusion Research Establishment, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities, during the period from 1 April, 2003 to 31 March, 2004. The activities in the Naka Fusion Research Establishment are highlighted by researches in JT-60 and JFT-2M, theoretical and analytical plasma researches, research and development of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction.

  In the JT-60 research program, the pulse length of the tokamak discharge was extended successfully to 65 s (formerly 15 s) in order to demonstrate/study the high performance tokamak plasma in the time scale comparable to/longer than the relaxation time of the plasma current profile. The control systems, the power supply systems, the heating systems and the diagnostic systems of the JT-60 were modified successfully to accomplish the extended discharge. The H-mode was extended successfully up to 24 s with high normalized beta value β_N~2.0 after the extension of the pulse length. The duration is more than twice of the world record in such high beta region. On the research of the advanced steady state tokamak, a high equivalent fusion gain Q_DT was achieved with full non-inductive plasma current. A large bootstrap current fraction ~75% (larger than the expected value in the ITER steady state operation) was sustained for more than 7 s in the negative shear plasma. Also the plasma operation was extended to high density and high radiation loss region using the high-field-side pellet injection and inert gas injection. On the research of the MHD instabilities and control, an early injection of the electron cyclotron (EC) wave was found to be effective for the stabilization of the neoclassical tearing mode (NTM). The transport of the energetic ions at the Abrupt Large amplitude Events (ALE) was studied. On the research of the H-mode physics, a non-dimensional pedestal identity experiment was carried out between JT-60 and JET in ELMy H-mode plasmas. On the research of the current drive, the experimentally obtained EC current drive efficiency was compared to a nonlinear Fokker-Planck code calculation with the effect of the toroidal electric field included. Initial results on the real time control of the safety factor profile by the lower hybrid (LH) waves were obtained. On the research of the divertor / SOL plasmas and plasma wall interaction, studies on the transient heat and particle load by the ELMs, SOL flow and impurity transport with the newly installed Tungsten tiles in the divertor progressed.

  The design of the new National Centralized Tokamak (NCT), which is the modification of the JT-60 to the super conductor version, progressed both in physics and in engineering utilizing the previous JT-60SC design. Two different configurations with low aspect ratio are studied.

  In the JFT-2M research program, the compatibility of the Ferrite inside wall (FIW) with high β_N plasma was demonstrated up to β_N~3.5. High β plasma was obtained with plasma even at the close wall case. In the study of the H-mode, an operational boundary of the high recycling steady (HRS) mode was identified, and the study of the characteristic oscillations in the edge transport barrier made a progress. Studies on divertor/SOL and compact toroid (CT) injection progressed.

  A series of the experimental programs on the JFT-2M was completed at the end of this fiscal year after the 21 years operations since 1983, with the significant contribution to the controlled nuclear fusion research.

  In the theoretical and analytical researches, significant progress was made in the studies of the H-mode power threshold, EC power necessary to stabilize the NTM, the effects of the ferromagnetic wall on the plasma stability and the effects of D shaping and rotation shear on the ballooning modes. In the project of numerical experiment of tokamak (NEXT), the studies of the ion temperature gradient driven trapped electron mode (ITG-TEM) turbulence, damping mechanism and global characteristics of the zonal flow 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 ITER superconducting magnets, the critical current density of a bronze processed Nb₃Sn strand has been improved by 14%, satisfying the specification of 700 A/mm² that was optimized for ITER. For development of ITER Neutral Beam Injector, H- ion beam of 110 mA (80 A/m²) was stably accelerated at the extraction voltage of > 5 kV. By the Cu coating on the bellows part of a 170-GHz gyrotron for ITER, the heating rate of the bellows was reduced to less than 1/10 of the original rate and 0.5-MW operation for 100 s was demonstrated successfully. For the Blanket fabrication technology, a joint technology between tungsten and F82H steel was developed using a solid state bonding method. The joining temperature and pressure were 960 C and 50 MPa, respectively in order to obtain fine grained F82H. To examine applicability of the screw tube for the plasma facing component of a fusion DEMO plant, thermal fatigue experiment of divertor mockups with a screw tube made of F82H has been performed at heat flux conditions of 3 and 5 MW/m². For the structural material development, irradiation experiments of F82H steel by multiple ion beams have been carried out up to 100 dpa to investigate the radiation hardening of F82H at high doses. The radiation hardening with increasing dose was almost saturated at around 30dpa. As an advanced method for tritium removal from ITER Test Blanket Module, research on the electrochemical hydrogen pump using a ceramic protonic conductor has been carried out extensively. D-T neutronics experiments of blanket mock-up assemblies designed by JAERI were performed. The mock-up consisted of lithium-6 enriched Li₂TiO₃, beryllium, F82H and a SS-reflector. The Tritium Production rates (TPR) measured and the TPR obtained by design analysis agreed each other within 10 %.

  In the ITER Program, along the work plan approved on February 2003 under the framework of the ITER Transitional Arrangements, the Design and R&D Task works started among the Participant Teams. JAERI has been in charge of seventeen Design Task works that make the implementation of preparing the procurement documents for facilities and equipments that are so scheduled as to be ordered at the early stage of ITER construction. The structure and management system and the staffing and resources of the ITER organization, the procurement allocation of ITER components and the site issue have been continuously discussed among the delegations of six countries/area including South Korea that joined the Negotiation in June 2003.

  Finally, in the fusion reactor design studies, the conceptual design of the economical and compact low aspect ratio (A~2) reactor (VECTOR) progressed. Researches on the physics related to the reactor design, liquid wall divertor and assessment of the fusion energy progressed.