The Fusion Engineering Research ProjectResearch Activities

The Fusion Engineering Research Project is intended to undertake both the detailed design of a steady-state fusion demonstration reactor and various engineering challenges to make it possible to construct the fusion reactor. The LHD-type reactor does not need any plasma current, and this feature gives the great advantage of realizing a steady-state reactor. It is essential to enhance engineering research based on this feature. The project is carrying out research on key components in fusion reactors, such as superconducting coil systems, high performance blankets, first wall, and divertors, while maintaining consistency with the demonstration reactor design. Serving as the center of fusion engineering research in Japan, this project enhances domestic and international cooperation for reactor design work and at the same time encourages basic research in the related interdisciplinary areas.

Superconducting Magnet System Research Group

Superconducting Magnet System Research Group

The large-scale magnet system for a fusion demonstration reactor requires high-performance superconductors with a 100 kA-class current capacity. Research is being conducted for developing such an advanced conductor made of metallic low-temperature superconducting materials such as Nb3Sn and Nb3Al and/or copper-oxide high-temperature superconducting materials such as YBCO. Actual environmental testing is also carried out to estimate the characteristics of superconducting materials under conditions at cryogenic temperatures, in an intense magnetic field and under neutron irradiation. Components in magnet systems are subjected to huge electromagnetic forces. Research is ongoing so as to precisely evaluate the expected stress on component materials and to seek the optimum coil supporting structure. The engineering design of the coil winding and fabrication method is also in progress.

The superconducting split-coil magnet facility that provides a  9-tesla magnetic field for testing superconductor samples with the maximum current  of 75kA.

The superconducting split-coil magnet facility that provides a 9-tesla magnetic field for testing superconductor samples with the maximum current of 75kA.

In-vessel Component Research Group

In-vessel components include the vacuum vessel, blanket system, first wall, and divertor. It is essential in blankets to use structural materials whose radioactivity decays rather rapidly after shutdown. Evaluations for a low-activation vanadium alloy have been carried out in collaboration with universities in Japan. In addition, experimental work to develop liquid breeder blankets using molten salts or liquid metals is being carried out on the improvement of material strength and control of corrosion at high temperature, which are important for the longevity and maintenance of a reactor. Furthermore, the first wall is required to serve as part of the blanket structure, facing the edge plasma, under 14.1 MeV neutron irradiation at elevated temperatures. In the project, the plasma-interactions with and hydrogen permeation through selected candidate materials are currently investigated. On the other hand, divertor heat flux in an LHD-type fusion demonstration reactor is considered to be 10MW/m2. Here highly heat-resistant divertor plates need to be developed. Three important subjects in the research and development are material selection, development of bonding technologies between armor tiles and their heat sink, and design studies of the 3D-shape of a helical divertor.

The high-temperature creep testing facility for small specimens
Samples of vanadium alloys can be tested at elevated temperatures up to 800°C in the ultra-high vacuum.

A mockup divertor tested in the heat load test device ACT (Active Cooling Test-stand), where the maximum electron-beam output is 100kW, the accelerated beam voltage is 30kV, and the maximum beam current is 3.3A).

A test sample in the plasma-wall interactions research facility is under plasma bombardment.

Reactor System Design Research Group

The development of a system design code and a neutronics code for selecting the primary specifications of a reactor is being conducted. The cost evaluation for power generation and the proposition of an operation schedule, taking into account unusual events, are also being conducted. The design optimization for the main components of a fusion power plant (plasma, superconducting magnet, blanket, and shield) is in progress.

Tritium is a radioactive isotope and therefore should be handled safely and with caution. This project includes the development of tritium handling and safety technologies, such as those for tritium decontamination and tritium removal. High sensitivity tritium detectors are also under development for environmental safety.

An example of a design window using the system design code. Design parameters including the magnetic field strength, the reactor size, the blanket thickness, the size of superconducting magnets, and core plasma parameters are comprehensively optimized.