Multiscale Stress Analysis and 3D Fitting Structure of Superconducting Coils for the Helical Fusion Reactor

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

Mechanical Design Concept of Superconducting Magnet System for Helical Fusion Reactor

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

Effect of coil configuration parameters on the mechanical behavior of the superconducting magnet system in the helical fusion reactor FFHR

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

Progress in the Conceptual Design of the Helical Fusion Reactor FFHR-d1

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

Bridge-Type Mechanical Lap Joint of a 100 kA-Class HTS Conductor having Stacks of GdBCO Tapes

In this paper, we reported design, fabrication and test of a prototype 100-kA-class high-temperature superconducting (HTS) conductor, especially for joint section, to be used for segmented HTS helical coils in the FFHR-d1 heliotron-type fusion reactor. The conductor has a geometry of three rows and fourteen layers of Gadolinium Barium Copper Oxide HTS (GdBCO) tapes embedded in copper and stainless steel jackets and has a joint section with bridge-type mechanical lap joint. We introduced improved method to fabricate the joint based on pilot experiments and we were able to apply a current of ∼ 120 kA at 4.2 K, 0.45 T to the sample without quench at joint. The obtained joint resistance was ∼ 2 nΩ, which was lower than our previous data. Though joint resistance increased with a rise in current and magnetic field, predicted joint resistance in the environment of actual helical coil in the FFHR-d1 was small enough to properly run the cryoplant of the reactor

Design modification of structural components for the helical fusion reactor FFHR-d1 with challenging options

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