Annual Report of Naka Fusion Research Establishment
From April 1, 2000 to March 31, 2001
Naka Fusion Research Establishment
Japan Atomic Energy Research Institute
Naka-machi, Naka-gun, Ibaraki-ken
(Received October 29, 2001)
This report provides an overview of research and development activities at Naka Fusion Research Establishment, JAERI, during the period from April 1, 2000 to March 31, 2001. The activities in the Naka Fusion Research Establishment are outstanding at high performance plasma researches in JT-60 and JFT-2M, and development in ITER EDA including technological R&Ds.
The JT-60 project aims at contributing to the physics R&D for ITER and establishing the physics basis for a steady state tokamak fusion reactor like SSTR. For the achievement of those objectives, both physical and engineering researches have been done. The JT-60 have continued to be productive in many areas covering performance improvements of high βp ELMy H-mode regime and reversed shear plasma, non-inductive current drive, physics study relevant to improved modes, stabilization of MHD modes, feedback control, disruption study, understandings on energetic particles, and scrape off layer and divertor studies with increased pumping capability. Highlights of FY 2000 experiments are summarized as follows:
1) In the high βp ELMy H-mode regime, discharges with a high total-performance have been sustained near the steady-state current profile solutions under full non-inductive current drive. In this experiment, the world record value of NB current drive efficiency ηCD of 1.55x1019 A/m2/W was achieved with increasing central electron temperature of 13keV.
2) The demonstration of the active control of ITB was performed by changing the direction of the toroidal momentum input. In this experiment, it is found that the change in Er shear of outer half region of the ITB layer (near the ITB foot) was important to control the whole ITB region.
3) Resistive instabilities are investigated in reversed shear discharges and it is found that the radially localized resistive interchange mode near the negative shear region leads to a major collapse through nonlinear mode coupling with a tearing mode in the positive shear region.
4) By increasing the input power of the fundamental O-mode EC wave, the complete stabilization of Neoclassical Tearing Mode was achieved. In a typical discharge where the complete stabilization was achieved, EC driven current density is calculated to be about 0.15kA/m2, which is about twice as large as bootstrap current density.
5) The deeper penetration for HFS(upper) pellet is consistent with the radial displacement based on the theoretically predicted E\B drift model. Using the HFS(upper) pellets, the accessible density region of the high βp plasmas is extended to 70% of the Greenwald density with H89P of 1.94.
6) By injecting Ar, the confinement of ELMy H-mode has been improved with high radiation loss power at high density. When the Ar density was higher than 0.5%, the HH-factor remained near unity in the range of ne < 0.65 nGW. The confinement improvement seemed to be closely related to the high ion temperature at the pedestal.
The highlights of technological progress in JT-60 are as follows;
1) A new guide tube for pellet injection from the midplane at the high magnetic field side was installed inside the vacuum vessel, and the pellets have been successfully injected with an minimal speed of 0.2 km/s.
2) A new boronization system for the plasma surface components using B10D14 with He as a carrier gas was developed, and in comparison with the former boronization using B10H14, the boronization time decreased to about one-fourth by stable glow discharge and the number of plasma discharges for wall conditioning after the boronization also decreased to one-tenth due to little hydrogen conent in the boron film.
3) A new real-time plasma shape reconstruction system based on the Cauchy-condition surface method was successfully applied to real time plasma equilibrium control for the first time in the world.
4) To test the pulsed operation of the ITER super-conducting center solenoid model coil, a control gain of the poloidal field coil power supply for JT-60U has been adjusted, and as the result the pulsed operation test completed successfully.
5) The development for increasing the beam power on the negative-ion based neutral beam injection system has been progressed, and as the result the power level of around 5MW at 400keV has been injected stably.
6) The fourth gyratron and transmission system for EC heating and ECCD were installed aiming at an injection power of more than 2 MW. The antenna of the new system can scan the EC beam in both toroidal and poloidal directions.
7) The detailed design of the JT-60 modification utilizing superconducting coils(JT-60SC), which aims at establishing scientific and technological bases for an advanced operation in an economically attractive DEMO reactor and ITER, has been completed.
On JFT-2M, advanced and basic research of tokamak plasma is being promoted, including application of the low activation ferritic steel, with the flexibility of a medium-sized device. The pre-testing on compatibility of ferritic steel plates (FPs), covering ~20% of the inside wall of the vacuum vessel, with plasma was performed, demonstrating no adverse effects on plasmas. Boronization was introduced for the first time in JFT-2M after installation of inside FPs. High-βN discharges (βN up to ~2.8) were obtained with inside FPs and boronization. Formation of negative electric field at the H-mode transition during ECH was clarified by the heavy ion beam probe (HIBP). The MSE polarimeter system, which is capable of simultaneous measurement of a radial electric field, has been newly developed. In RF experiments, fast wave electric field profile was directly measured for the first time using the beat wave and HIBP.
The principal objective of theoretical and analytical studies is to understand physics of tokamak plasmas. 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.
The optimum condition of electron cyclotron (EC) beam injection has been efficiently obtained to stabilize neo-classical tearing modes (NTM). Remarkable progress has been made in the study of turbulence driven by electron temperature gradient instabilities. The study on the formation of a detached-plasma has been much progressed.
Major items of Research and Development (R&D) of nuclear fusion reactor technologies, mainly focused on ITER-related areas in FY2000 are as follows:
1) Blanket: Be/DSCu specimens made by HIP method with an Al interlayer have successfully withstood against a heat flux of 5 MW/m2 for more than 1000 cycles.
2) Superconducting Magnet: The world’s largest superconducting pulsed coil was successfully tested at the target field of 13 T and operating current of 46kA, with a stored energy of 640 MJ at a ramping rate of 1.2 T/s.
3) NegativeIon Beam: 1MV voltage folding of the beam accelerator column was successfully demonstrated. The negative ion production mechanism was clarified in detail.
4) RF Heating: A 170GHz gyrotron was successfully tested at 1MW level power for about 10 second (0.9MWx9.2sec).
5) Tritium handling: Performance of ITER-scale 2,500 m3/hr large atmosphere detritiation system was successfully confirmed. A new separation method could separate the mixture gas of H2/He to each composition gas with the enrichment more than 99%.
6) Plasma Facing Components of Divertor: A prototype mockup of the ITER divertor could successfully sustain a heat flux up to 20MW/m2 (15sec) for more than 1000 thermal cycles.
7) Reactor Structure: The blanket module has been replaced under the required clearance of ±0.25 mm between key and groove by remote handling.
8) Fusion Neutronics: Radiation detectors using single crystal CVD diamonds have been developed aiming the 14 MeV neutron spectrometer. A key element technology phase of IFMIF has begun to reduce the key risk factor for its construction.
An option of minimum cost still satisfying the overall programmatic objective of the ITER has been developed by the three Parties of Japan, EU and RF during the extension of the Engineering Design Activities (EDA) after the successful completion of the six year EDA in July 1998. The Outline Design Report of the newly developed ITER-FEAT was submitted in January 2000 for the review of the Parties. The technology which had been established through the plasma physics and technology R&D during the past EDA or the technology being established by the current R&D were employed in the design of ITER-FEAT. The ITER Council subsequently approved the design in June 2000 as a single mature design for ITER consistent with its revised objectives. The Draft Final Design Report of ITER-FEAT was prepared for technical review by the Technical Advisory Committee (TAC) held in February 2001. The TAC concluded that the ITER-FEAT was ready for a decision on construction with recognition of the achievement in reducing the cost by 50%.
In April 2000, the successful DC operation of the CS Model Coil which had been developed since the beginning of the EDA was achieved in JAERI Naka. In August 2000, ten thousand cycles of pulsed operations were achieved to simulate ITER full-scale CS. The full-scale sector of the Vacuum Vessel sector with port extension for testing of the remote welding and cutting had been dismantled in March 2001 after successful completion of the test.
In fusion reactor design, the physics design of a fusion power reactor A-SSTR2 was developed on plasma current startup. At the same time, hydrogen production from biomass and other various uses of high-temperature helium gas coolant from the A-SSTR2 plant were proposed. As to safety research on A-SSTR2, a reactor design concept to reduce radioactive wastes after the decommissioning is proposed.
|Editors||:||Kuriyama M., Kizu K., Kusakawa F., Matsumoto H., Sakamoto K., Sengoku S.|
|Keywords||;||JAERI, Fusion Research, JT-60, JFT-2M, NEXT, Fusion Engineering, ITER, EDA, Fusion Reactor|