Mechanical Design Concept of Superconducting Magnet System for Helical Fusion Reactor

Hitoshi Tamura,Nagato Yanagi,Takuya Goto,Junichi Miyazawa,Teruya Tanaka,Akio Sagara,Satoshi Ito,Hidetoshi Hashizume

doi:10.1080/15361055.2019.1603041

Hitoshi Tamura, Nagato Yanagi + Show 6 more

Open Access

https://doi.org/10.1080/15361055.2019.1603041

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Abstract

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.

Highlights

Helical fusion reactors have attractive features, such as steady-state operation in the absence of a plasma current drive and a built-in helical diverter
The National Institute for Fusion Science is developing a conceptual design of the Large Helical Device (LHD) [1]- type helical reactor, FFHR [2, 3]
In the structural analysis of FFHR-c1, a gas-cooled high-temperature superconductor (HTS) with Rare-earth barium copper oxide (REBCO) coated superconductor was adopted since the maximum magnetic field on the coil was estimated to reach 19 T, and to be able to operate at high temperatures

Summary

Introduction

Helical fusion reactors have attractive features, such as steady-state operation in the absence of a plasma current drive and a built-in helical diverter. The National Institute for Fusion Science is developing a conceptual design of the Large Helical Device (LHD) [1]- type helical reactor, FFHR [2, 3]. The original concept was based on force-free-like configuration of helical coils (HCs) to reduce electromagnetic (EM). The geometries of the magnet systems for both types of FFHR, including a pair of HCs and two sets of vertical field coils (VFCs), are based on the geometry of LHD. LHD has a major radius of 3.9 m and magnetic field intensity of 4 T (in design with 1.8 K superfluid helium cooling: actual achievement is 3 T with 4 K liquid helium). Weight estimation with the virial theorem and prospects for future fusion reactors are discussed as well

Structural design conception

Weight estimation

Findings

Summary