Akio Sagara

While a tokamak has a plasma current disruption, a helical system is much more robust because the current-less operation is, in principle, possible. However, a bootstrap (BS) current due to the pressure gradient could be a concern because it could change the rotational transform. We conducted experiments on the BS current in the Large Helical Device (LHD) in the 1/ ν and plateau regimes using three magnetic axis configurations R _ax = 3.75 m, 3.6 m and 3.53 m with completely balanced neutral beam injection (NBI). As a reference, we use the standard configuration of R _ax = 3.75 m. We then examine the inward-shifted configuration of R _ax = 3.6 m, and the strongly inward-shifted configuration of R _ax = 3.53 m, which is referred to the neoclassically optimized configuration (Murakami et al 2002 J. Plasma Fusion Res . 5 620;, Murakami et al 2002 Nucl. Fusion 42 L19). It was demonstrated that the BS current for R _ax = 3.53 m is clearly lower than that for R _ax = 3.75 m, together with a confinement improvement. Based on these experimental results, the BS current coefficient has been derived, which helps to predict the BS current in a helical fusion reactors.

A novel method, called Advanced Multi-Step Brazing, was developed to fabricate a new type of divertor heat removal component with W armor and an oxide-dispersion-strengthened copper (GlidCop®) heat sink in the initial phase of our work. Later, a new type of divertor heat removal component, which has a rectangular-shaped cooling channel with a V-shaped staggered-rib structure in the GlidCop heat sink, was developed. This new component showed an extremely high heat removal capability during a ~30 MW/m² steady-state heat loading condition in our previous work. In this work, the new component was installed in the divertor strike position of the Large Helical Device and exposed to neutral beam injection–heated plasma discharges with 1180 shots (~8000 s) in total. Though submillimeter-scale damage, such as unipolar arc trails and microscale cracks, was identified on the W surface, the extremely high heat removal capability did not show any sign of degradation over the experimental period. On the other hand, remarkable sputtering erosion and redeposition phenomena, due to the strong influx of the divertor plasma, was confirmed on the W armor.

We are proposing to connect a supercritical CO₂ gas turbine (SCOT) power generator to the Operational Recovery of Separated Hydrogen and Heat Inquiry-2 (Oroshhi-2) nuclear fusion primary molten salt test loop constructed at the National Institute for Fusion Science and conduct a demonstration test of the power generation principle from the world’s first nuclear fusion simulated heat output. In this paper, we describe the results of a design study of the FLiNaK/CO₂ intermediate heat exchanger and SCOT power generation system and related equipment that can be connected to the molten salt test loop. From the results of this study, it was concluded that the power generation principle verification test with 20-kW(electric)–class power generation with an efficiency of 20% is possible by the SCOT system connected to the FLiNaK test loop in Oroshhi-2 equipment via the intermediate heat exchanger.

This study evaluates convective and boiling heat transfer characteristics of a water impinging jet flow in porous media in order to remove the heat flux of 10 MW/m² imposed to fusion divertors. The metal porous media with complicated microchannels have large heat transfer surface due to fin effect and superior mixing effect of fluid, which enhances not only the convective heat transfer but also the boiling heat transfer by improving the evaporation rate of the cooling liquid. In a proposed heat removal device called EVAPORON-3-Type3, the cooling water is supplied as an impinging jet flow into the porous medium, which is a two-layered copper particle bed, and the generated vapor is discharged through high porosity gaps on the heat transfer surface. As a result, the convective heat transfer coefficient is improved by 1.6 times compared with that of an impinging jet flow without the copper particle bed. In the boiling heat transfer regime, the critical heat flux is increased by 1.5 times and the heat flux of 8.4 MW/m² is achieved under low velocity and highly subcooled conditions though it’s not maximum.

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.

One of the key issues in the design of LHD-type heliotron reactors is to secure sufficient blanket spaces. In this respect, helical pitch parameter γ is quite important because it significantly affects both the coil and plasma shapes. For a quantitative understanding of the effect of helical pitch parameter on a design window, a system design code for LHD-type heliotron reactors HELIOSCOPE (HELIOtron System design COde for reactor Performance Evaluation) is developed and parametric scans are carried out with three cases of γ = 1.15, 1.20 and 1.25. It became clear that design windows of heliotron reactors depend significantly on the engineering constraints: the stored magnetic energy of the coil system, the inboard minimum blanket space and the averaged neutron wall load. In the case of a fusion power of 3 GW, γ = 1.20 is optimum for relaxing physics requirements. But γ = 1.15 is also a possible selection if a lower fusion power or a higher neutron wall load is accepted. Since design windows are quite sensitive to the engineering constraints and physics conditions, a further detailed study on design feasibility of advanced engineering components and the effect of γ on the physics conditions is expected to optimize the value of γ.