Nuclear Fusion is an excellent energy source characterized by high safety and inexhaustible fuel. Now fusion research is coming to the stage of ‘system integration.’ Towards demonstration of the burning plasma and integrated fusion engineering, construction of ITER is underway in St Paul Lez Durance, France. In order to contribute to the success of ITER and complement ITER towards the demonstration reactor (DEMO), Japan and Europe are operating ‘The Broader Approach Activity’ in Japan, in which the JT-60SA tokamak device is under construction in Naka city, Ibaraki prefecture for development of steady-state high pressure plasma operation and the IFMIF prototype accelerator for the fusion neutron engineering is under construction in Rokkasho village in Aomori prefecture. Fostering the next generation is also the important purpose of these activities. “The All Japan Activity” was recently organized for DEMO design and development.
The framework of industry-university-government cooperation has been organized to establish technology bases required for the development of a fusion DEMO reactor, and discussion on the roadmap and action plan for realization of the fusion energy has been actively launched. The target and subjects on fusion DEMO reactor are expected to be identified based on the essence of energy policy. Furthermore, the Joint Special Design Team for fusion DEMO has been established, and has been developing the concept of Japan’s fusion DEMO reactor.
In the ITER project at present, construction of the buildings is progressing at the site of Saint Paul les Durance, South France. And in parallel, in the seven participating Parties, Japan, Europe, the USA, Russia, China, Korea and India, the design and manufacturing of the ITER core equipment is progressing. In Japan, manufacturing of the real ITER components, such as Toroidal Field (TF) and Central Solenoid (CS) conductors, TF coil and its structures, 1 MV high voltage power supplies for Neutral Beam heating system and gyrotron for Electron Cyclotron heating system, are in progress. In November 2016, the ITER Council, consisting of high-level government officials, approved a new overall project schedule in which the ITER’s first plasma is scheduled in December 2025, followed by a staged expansion of the operation regime to start nuclear fusion reactions in 2035.
JT-60SA is the superconducting large tokamak device under construction by Japan and EU in Naka, Ibaraki, Japan. The project mission of JT-60SA is i) to develop the steady-state operation at high plasma pressure required for DEMO reactors, which cannot be done in ITER, and ii) to support ITER with various new ideas. Fostering the next generation scientists for ITER and DEMO is also the main project mission. Fabrication and installation of components and systems are steadily progressing with high accuracy of manufacture and assembly towards start of operation in September 2020. The JT-60SA Research Plan has been issued by 378 co-authors from 45 institutes 15 countries in Japan and EU.
Development of an IFMIF/EVEDA prototype accelerator aimed at a high intensity neutron source is ongoing. In the initial stage injector test, the target 100 keV–140 mA deuteron acceleration is demonstrated. This beam is accelerated to 5 MeV by the subsequent high-frequency quadrupole linear accelerator and is planned to be accelerated to 9 MeV with a superconducting accelerator. The EVEDA Lithium Test Loop which is the target of the deuteron beam for IFMIF demonstrated liquid lithium flow of 15 m/s, stable operation for 1,300 h.
Based on the Broader Approach agreement, IFERC project implements three sub-projects; DEMO Design and R&D Coordination Centre, Computational Simulation Centre and ITER Remote Experimentation Centre, in order to complement ITER and contribute to an early realization of DEMO reactor. Herein, activities of the three sub-projects are outlined.
The integrated performance required for the fusion plasma is to produce sufficient fusion power by high energy confinement, to keep the heat flux onto the first wall within the allowable level, to make the reactor core compact by high fusion power density (=high plasma pressure), and to sustain the plasma in the steady-state under a small circulating power in the power plant. Japan has been leading the fusion plasma research in the world by achieving various records such as the break-even condition in JT-60 etc. The key research areas towards DEMO reactors include i) achievement and long sustainment of the burning plasmas (the energy gain=10) in ITER, ii) achievement and long sustainment of steady-state high pressure plasmas (high normalized pressure and high fraction of self-driven plasma current), and iii) understanding of the fusion plasma together with modeling and simulation enabling sufficient prediction of the DEMO reactors.
A superconducting magnet system is one of major components in a fusion reactor to generate the magnetic field for the plasma confinement. National Institutes for Quantum and Radiological Science and Technology (QST) has completed 25％ of Toroidal Field (TF) conductors in the procurement of the ITER magnet system. All Central Solenoid (CS) conductors, 9 TF coils and 19 TF coil cases are currently under manufacturing.
Recent progress to realize the neutral beam (NB) systems for ITER and JT-60SA is reported. In the negative ion source as the primary beam source of the NB system, a vacuum breakdown between acceleration grids and excess power loading on the acceleration grid due to beam particles have limited to demonstrate the MeV-class and long pulse beam accelerations. To solve these issues, a vacuum insulation with large acceleration grid were established and precise control technique of the beam particles have been developed. As the results, 1 MeV beams with the relevant current density of the ITER requirement has been successfully demonstrated for 60 s seconds stably. In the 1MV dc high voltage power supply components for the ITER NB test facility (NBTF) under construction in Italy, all components have been successfully completed based on the R&Ds. As the results, 90％ of them have been installed in the NBTF site in 2017 and the integrated test of the power supply will be completed in 2018. Finally, the overview for the DEMO NBI based on these R&Ds is described.
Radio frequency heating systems are essential for nuclear fusion research to heat the plasma or drive a current by accelerating electrons or ions confined magnetically. As for the electron cyclotron heating system, a go-ahead method among them, development shows steady progress. Especially the performances of electron tube “gyrotron” which generates a megawatt of radiofrequency power, have already meet the requirement of ITER and JT-60SA, leading the world. The procurement of the electron cyclotron heating system has already been started to these tokamaks.
In large tokamak device, high current DC power supply several tens of kA/several kV has been used for magnet excitation. During tokamak operation, high voltage is needed for only short time at plasma initiation. Therefore, cost-effective circuit topology combined with thyristor converters and DC circuit breakers has been discussed for many years. In addition, the latest superconducting tokamak stores huge magnetic energy(～10 GJ) especially in toroidal field coils. In the case of coil quench, protection circuit has to discharge the magnetic energy by using external resistors in order to prevent coil damage. This paper describes an optimized magnet power supply system of JT-60SA which is one of the largest superconducting tokamak and being constructed in Japan.
In nuclear fusion reactor, Tritium Breeder Blanket has three functions such as the shielding of fusion neutron, the extraction of high quality heat for electrical power generation, and the production of fuel tritium. So, the blanket is a key equipment and technology for fusion power plant. In this paper, the outline of the tritium breeder blanket is mentioned. And TBM program implemented by Japan Domestic Agency within the ITER project and its current status are also introduced.
Fusion material development is the key R&D issues to realize the fusion blanket system which is expected to transform fusion energy to thermal energy and breed tritium under high energy fusion neutron bombardment. The structural material is expected to endure not only variation of mechanical loads but also high heat load and high energy neutron damage, and functional materials such as tritium breeder and neutron multiplier are expected to be chemically stable during the service. The current status of these fusion materials are summarized.
Recently engineering researches on a plasma facing component under highest heat flux in a nuclear fusion reactor have been developed with R&Ds on ITER and JT-60SA divertor. Especially, this paper reports achieved significant milestones which are not only durability for heat loading of full-Tungsten divertor but also accuracy surface profile after assembly of the plasma facing components.
Nuclear fusion research has entered the demonstration stage of the reactor engineering technology, and the design activity of the ITER tritium system for deuterium (D)/tritium (T) fusion reaction is proceeding. Tritium confinement in the facility and detritiation with the detritiation system (DS) are the keys to secure the safety of the fusion reactor. Japan Domestic Agency has established the DS joint procurement team with the ITER organization and conducts DS final design activity and DS qualification tests.
Remote maintenance systems are inevitable in fusion reactors to perform maintenance in the vacuum vessel, which must be high-level radiation environment. National Institutes for Quantum and Radiological Science and Technology is procuring the ITER Blanket Remote Handling System and carrying out R&D activities on unachieved remote-maintenance technologies such as pipe welding.
Plasma diagnostic systems make penetrations through plasma facing components, shielding blankets, safety boundary, fire sector and ventilation zoning, and hence, it is required to demonstrate its feasibility of nuclear safety. ITER is a first of a kind nuclear fusion reactor, in which nuclear safety design shall be built-in even in the diagnostic systems. This paper describes what the authors have found in the design activity of ITER poloidal polarimeter that Japan procures. The knowledge can be extrapolated for designing plasma diagnostic systems in future nuclear reactors such as prototype reactors.
For more accurately evaluation of the damage caused in the blanket structural materials in the nuclear fusion reactor and to verify their soundness during the service, under irradiation by the fusion neutron at a level that the ratio (He/dpa) of the helium production rate and displacement rate becomes close to 10, it is necessary to verify the material characteristics change due to the interaction between the defect created and the helium generation by various material tests. Therefore, verification using a high intensity neutron source with neutron spectrum simulating the fusion reactor like IFMIF is essential. In this chapter, based on the development technologies by IFMIF/EVEDA, a project for the construction of Advanced Fusion Neutron Source (A-FNS) facility planned by QST with proposals for its usage for various purposes is introduced.