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The conceptual design of a helical fusion reactor was studied at the National Institute for Fusion Science in collaboration with other universities. Two types of the force free helical reactor (FFHR) are FFHR-d1 and FFHR-c1. FFHR-d1 is a self-ignition demonstration reactor that operates with a major radius of 15.6 m at a magnetic field intensity of 4.7 T. FFHR-c1 is a compact subignition reactor that aims to realize steady electrical self-sufficiency. Compared to FFHR-d1, FFHR-c1 has a magnetic field intensity of 7.3 T and a geometrical scale of 0.7. The location of the superconducting coils in both types of FFHR is based on that of the Large Helical Device (LHD). LHD has a major radius of 3.9 m. According to the design of LHD, the deformation must be within the required value to compensate for the accuracy of the magnetic field. According to this concept, the magnet support structure of LHD was fabricated using thick Type 316 stainless steel to impart sufficient rigidity. Thus, the stress of the magnet system of LHD is sufficiently below the permissible stress. In the case of FFHR, from the viewpoint of the reactor, a large access port is required for the maintenance of the in-vessel components. The mechanical design of the support structure is conceptualized by considering the basic thickness of the material and residual aperture space by referencing the mechanical analysis results. Details of the design concepts of LHD and FFHR-d1/FFHR-c1 as well as the results of mechanical analyses are introduced in this paper.

FFHR-d1A and c1 are the conceptual design of a helical fusion reactor. The positional relationship among superconducting coils, a pair of helical coils with two sets of vertical-field coils, are observed to be similar in both type of FFHR. Such a relation of coil configuration is based on the coil configuration of the Large Helical Device, which has been designed and constructed at the National Institute for Fusion Science. There is increasing demand to achieve an optimized coil configuration to anticipate improvements in plasma-confinement conditions. In this study, the structural design of FFHR based on the fundamental set of parameters of coil configuration is depicted, which satisfies the soundness of the structure. Further, the effects of the coil configuration parameters on the stress distributions are investigated. An effect of radius of curvature on a winding scheme of the helical coil is also discussed.

A High-Temperature Superconducting (HTS) magnet is being considered to for use in the next-generation helical experimental devices. Three types of large-current HTS conductors are being developed, and one of them is the STARS (Stacked Tapes Assembled in Rigid Structure) conductor which uses HTS tapes with a simple stacking technique. Following the proof-of-principle experimental results obtained in the former 100-kA-class prototype hand-made conductor sample, an actually applicable conductor is being developed with a rated current of 18 kA at a temperature of 20 K and a magnetic field of ∼10 T. One of the crucial requirements for this conductor is to have a high current density of 80 A/mm2. In the first phase of the development, a 3-m short sample was fabricated by applying laser-beam welding to the stainless-steel jacket. It was tested in liquid nitrogen at 77 K with no external magnetic field. Then the sample was tested in gaseous helium at 20 - 40 K under a magnetic field of 6 - 8 T, and the results show that the basic requirements were satisfied.

High-temperature superconducting (HTS) conductor is a feasible candidate to make magnets for the next generation fusion devices because of its higher temperature margins and higher critical current in a high magnetic field in comparison to low-temperature superconducting (LTS) conductors. The recently proposed concept of the HTS-WISE (Wound and Impregnated Stacked Elastic tapes) conductor was studied to clarify its characteristics under certain magnetic fields. The WISE conductor, including 30-stacked REBCO (Rare-Earth Barium Copper Oxide) tapes, was fabricated and energized in a 9-T test facility which produced the condition of magnetic field B = 5 - 8 T and a temperature T = 30 - 50 K. Obtained critical currents (5.4 - 10.8 kA) increased with a decreasing magnetic field and/or temperature under the condition of T > 40 K. The maximum current of 16.9 kA was obtained at T = 30 K, which corresponded to the engineering current density jE = 60 A/mm2. Experimental results showed qualitative agreement with numerical calculations of the critical current. We confirmed the operation of the WISE conductor under a high magnetic field and low temperature.

The LHD-type helical fusion reactor FFHR has been studied to realize steady-state fusion power generation without a need for current drive and free from disruption. The conceptual design studies of FFHR are steadfastly progressing based on the presently ongoing experiments in the Large Helical Device (LHD). In order to enhance the attractive features of the base option of FFHR-d1A, which is similar to LHD, configuration optimization is being considered for FFHR-d1C. Slight modification of the helical coil trajectory gives an improved condition both for the plasma confinement and the MHD stability. In order to overcome the difficulty for construction and maintenance associated with the three-dimensional structure, innovative ideas are being explored for the superconducting magnet, divertor, and blanket. For the superconducting helical coils, the joint-winding method confirms a fast manufacturing process. The helical divertor is reexamined and practical feasibility is discussed. The maintenance method of the helical divertor and the helically-segmented breeder blanket is a serious issue and a plausible solution is proposed.

A conceptual design for a helical fusion reactor is currently being undertaken by the National Institute for Fusion Science, Japan. The coil support structure is designed from the perspective of the allowable stress of the material. A continuous helical coil winding with a low temperature superconductor and a water-cooled divertor made of tungsten and copper alloy have been considered for use in this reactor, and it is defined as the basic option. Several flexible design options have also been proposed; these options solve existing issues in the basic option and they are treated as the challenging options. They can be implemented by modifying the structural components of the basic option. The structural design modifications need to be made in order for the challenging options and the mechanical soundness were investigated.

An innovative idea is proposed for enlarging the blanket space on the inboard side of the torus for the helical fusion reactor FFHR-d1. A set of sub-helical coils, named NITA coils, with opposite-directed current outside the main helical coils, effectively reduces the helical pitch parameter and enlarges the blanket space. Dependence of the blanket space and plasma volume on the effective helical pitch parameter is examined. The obtained magnetic surfaces and their properties are compared with that of the original configuration.

Conceptual design studies for the Large-Helical-Device-type helical reactor, i.e., FFHR-d1, are being conducted in the National Institute for Fusion Science. Three different cooling schemes and conductor types have been proposed for the superconducting magnet system. A multiscale structural analysis is used to assess the mechanical characteristics of the magnet structure, taking into account the types of cooling schemes and superconductors. Multiscale analysis evaluates both the stress distribution in the coil support structure and local stress in the constituents of the superconductors without rebuilding a finite-element model of the support structure. Concerning a segmented fabrication of the helical coils using a high-temperature superconductor, the feasibility of segment installation is confirmed using a three-dimensional printing model, which identifies the maximum segment length and the necessary gap in the coil casing to install a segment.

Recent progress in plasma performance and the understanding of the related physics in the Large Helical Device is overviewed. The volume-averaged beta value is increased with an increase in the neutral beam injection (NBI) heating power, and it reached 5.0% of the reactor-relevant value. In high-β plasmas, the plasma aspect ratio should be controlled so that the Shafranov shift would be reduced, mainly to suppress transport degradation and the deterioration of the NBI heating efficiency. The operational regime of a high-density plasma with an internal diffusion barrier (IDB) has been extended, and the IDB, which was originally found using the local island divertor, has been realized in the helical divertor configuration. The central density was recorded as high as 1 × 1021 m-3, and the central pressure reached 130 kPa. Based on these high-density plasmas with the IDB, a new ignition scenario has been proposed. This should be a scenario specific to the helical fusion reactor, in which the helical ripple transport would be mitigated. A low-energy positive-NBI system was newly installed for an increase in the direct ion heating power. As a result, the ion temperature (Ti) exceeded 5.2 keV at a density of 1.2 × 1019 m-3 in a hydrogen plasma. Transport analysis shows improvement of ion transport, and the Ti-increase tends to be accompanied by a large toroidal rotation velocity of the order of 50 km/s in the core region. The plasma properties in the extended operational regime are discussed from the perspective of a steady-state helical fusion reactor.