Annual Report of Naka Fusion Research Establishment
From April 1, 1999 to March 31, 2000
Naka Fusion Research Establishment
Japan Atomic Energy Research Institute
Naka-machi, Naka-gun, Ibaraki-ken
(Received October 25, 2000)
This report provides an overview of research and development activities at the Naka Fusion Research Establishment, JAERI, during the period from April 1, 1999 to March 31, 2000. The activities in the Naka Fusion Research Establishment are highlighted by high performance plasma research in JT-60 and JFT-2M, and progress in ITER EDA, including ITER technology R&D.
Objectives of the JT-60 project are to contribute to the physics R&D for International Thermonuclear Experimental Reactor (ITER) and to establish the physics basis for a steady state tokamak fusion reactor like SSTR. For the achievement of these objectives, some mechanical improvement of the centrifugal pellet injector have been done for stable production and successive ejection of pellet. In addition, a guide tube for an injection from top of high-field side was installed. As a result, pellets were successfully injected from the top of high-field side as well as the low-field side in February 2000. The injection power of electron cyclotron heating (ECH) system of the frequency of 110 GHz installed last year was increased up to ~0.75 MW for 2 seconds in this year using the control of RF beam angle. Two gyrotrons were newly installed with total power increased from 1 MW to 3 MW.
The JT-60 experiments have continued to be productive in many areas covering performance improvement of reversed shear plasmas, non-inductive current drive, physics study relevant to improved modes, stabilization of MHD modes, feedback control, disruption study, understandings on energetic particles and divertor studies with increased pumping capability. Highlights of FY 2000 experiments may be summarized as follows:
(1) A reversed shear discharge with an equivalent fusion multiplication factor QDTeq of ~0.5 was achieved successfully at plasma current of 2.4 MA for 0.8 seconds.
(2) Quasi-steady operation of a low current reversed shear plasma with a large fraction (~80%) of bootstrap current was realized under full non-inductive current drive condition. H-factor of 3.3-3.8 at electron density as high as 73% of the Greenwald limit was sustained for 6 times the energy confinement time.
(3) Normalized beta exceeding the ideal no-wall stability limit was obtained in reversed shear plasmas with a ratio of an outer-wall radius to a plasma minor radius less than 1.3.
(4) L-H transition power was reduced by ~30% in the W-shaped divertor with pumping from both inside and outside slots compared with that in the open divertor. Helium exhaust rate in ELMy H-mode plasmas was improved up to 50% higher than the inside slot pumping.
(5) Current drive efficiency of 1.3´1019 A/m2/W was attained by N-NBI at 350 keV in the high bp ELMy H-mode with the central electron temperature of 8.6 keV. The efficiency is about 2.6 times higher than that of 100 keV.
On JFT-2M, advanced and basic research of tokamak plasma is being proceeded including the application of the low activation ferritic steel. A dramatic reduction of trapped ion loss due to the toroidal field ripple was identified for the first time in the tokamak experiment through measurements of the first wall temperature. Ferritic steel boards covering 20 % of the first walls have been installed inside the JFT-2M vacuum vessel for the coming experiments. Fast sampled measurement of potential and fluctuations by the heavy ion beam probe revealed characteristic time scales of the potential change at the plasma periphery for the first time at the L/H transition. The results would provide an important evidence for the L/H transition theory based on the change of electric field structure.
The principal objective of theoretical and analytical studies is to understand physics of tokamak plasmas. Much progress was made on analyzing dynamics of the internal transport barrier in JT-60U reverse shear plasmas. Progress was also made on the study of micro and macro instabilities. The NEXT (Numerical EXperiment of Tokamak) project has been progressed in order to investigate complex physical processes in core plasmas, such as transport and MHD, and in divertor plasma by using recently advanced computer resources. Remarkable progress was made on the development of divertor simulation codes.
R&D of fusion reactor technology has focused on the ITER EDA and DEMO-related areas including application to the other field. Major highlights in FY 1999 are as follows:
(1) CS Insert Coil which is one layer solenoid with Nb3Sn conductor was successfully fabricated and installed with inner and outer CS Model Coil modules into the CS Model Coil test facility.
(2) Thermal fatigue test of the first wall panel made by reduced activation ferritic steel F82H has been performed and the first wall panel withstood 8500 cyclic heating by 5.5 MW/m2 surface heat flux (surface temperature estimates 1000¼C).
(3) A new experiment to design the vacuum insulated beam source for the negative ion-based neutral beam injector (N-NBI) started. High-energy (725 keV) H- ions, which was produced by the ion source for N-NBI, were implanted into the single-crystal Si wafer without mass separation to delaminate a thin Si layer. After annealing the implanted Si wafer at 600¼C for 10 minutes, the Si layer has been successfully delaminated at a thickness of 10 mm.
(4) Critical heat flux test on screw tube under one sided heating conditions, thermal fatigue tests with 20 MW/m2 on short dump target, and disruption erosion tests on various tungsten were performed.
(5) Design study of International Fusion Material Irradiation Facility(IFMIF) responding to the deuterium beam current upgrade in three stages(50/125/250 mA), which corresponds with the fusion reactor program, has been progressed.
After the successful completion of the six year Engineering Design Activities (EDA) for ITER in July 1998, the three Parties of Japan, the European Union (EU) and the Russian Federation (RF) agreed to extend the period of the EDA for three years until July 2001 in order to modify the design of ITER so that the construction cost can be reduced as much as about a half of the cost of 1998 design, by taking account of the progress in plasma physics and technology R&D. For the design of the modified ITER (termed 'ITER-FEAT'), new technical guidelines were established, and various design options were examined in line with the guidelines. The concept of ITER-FEAT was eventually selected at the Program Directors' Meeting held in July 1999, with emphasis on high plasma density and the capability of steady-state operation. On the basis of this decision, the design of ITER-FEAT was developed, and the Outline Design Report together with its Technical Basis document were submitted by the ITER Director to the ITER Meeting of the Parties' representatives held in January 2000. The Technical Subcommittee of the Fusion Council of Japan reviewed these documents, and concluded in March 2000 that the predicted plasma performance and engineering performance are appropriate for fulfilling the technical objectives. Further development of detailed design of ITER-FEAT will be continued aiming at completing the Final Design Report, a full set of other technical documents, and ITER technology R&D by July 2001.
Based on the previous design studies of SSTR in 1990, advanced SSTR in 1996 and a concept of DREAM reactor with a high availability and an easy maintenance system, a new tokamak fusion power reactor (A-SSTR2) which meets both economical and environmental requirements was developed in 1999. Radiological toxic hazard potential is compared among all the radioactive materials contained in a fusion reactor, a pressurized light water reactor and a coal-fired power plant.
|Editors||:||Ninomiya H., Inabe T., Kaneko T., Konoshima S., Miura M. Y., Nakamura K.|
|Keywords||;||Fusion Research, JAERI, JT-60, JFT-2M, NEXT, Fusion Engineering, CS Model Coil, N-NBI, ITER, Fusion Reactor Design|