Helical Fusion Reactor Research Articles

Development of steady-state fusion reactor by Helical Fusion

The Helical Fusion group (Helical Fusion Co., Ltd. in Japan and Helical Fusion USA, Inc.) is developing a 50 MWe-class steady-state helical fusion reactor, which is a kind of stellarator called heliotron composed of two continuous helical coils similar to the large helical device and is operated without plasma current. HESTIA plays the role of the fusion pilot plant. The first-of-a-kind fusion power plant that would follow HESTIA will be a 100 MWe-class steady-state helical fusion reactor. After more than one year of continuous operation, maintenance will be completed within 3 months, aiming for an availability larger than 80%. High-temperature superconducting magnet coils are adopted to generate 8 T of the magnetic field at the helical coil center. HESTIA is a deuterium-tritium fusion reactor, where tritium is self-produced using liquid metal blanket systems. The first wall of the liquid metal blanket including the divertor strike zone is covered and protected by liquid metal free-surface flow, and therefore, individual divertor systems are not required in HESTIA. Electron cyclotron heating is adopted for plasma heating. Since plasma current drive is not required, HESTIA can be operated at a low fusion gain of ∼13, and steady-state operation is principally possible on the order of a year. After a few years of individual development phases, a prototype device is planned to be constructed and operated for the integrated demonstration before starting the construction of HESTIA.

Toward the realization of a commercial Helical Small-Scale Steady-State Fusion Reactor

Toward the realization of a commercial Helical Small-Scale Steady-State Fusion Reactor

Numerical Optimization of Electromagnet Current Distribution in Superconducting Linear Acceleration System

An enhancement of the acceleration performance of a superconducting linear accelerator when injecting pellets into a helical fusion reactor has been investigated numerically. For this purpose, a numerical code using the finite element method was developed to analyze both the time evolution of a shielding current density in a high-temperature superconductor film and the dynamic motion of the film. In the present study, to optimize the shape of double magnets, a multi-objective optimization problem is solved using the non-dominated sorting genetic Algorithm-II. The results of the computations show that the optimization of the inner electromagnet is more effective in improving the pellet velocity than that of the outer electromagnet. In addition, we found that the pellet speed and energy efficiency have a trade-off relation. Therefore, it is possible to determine the preferred solution for the design of an acceleration magnet, whether focusing on the pellet speed or the energy efficiency. In other words, this relationship can be used to choose whether to emphasize velocity or efficiency in the design of SLA systems.

Low-resistance joint development for segment-fabrication of high-temperature superconducting fusion magnets

Two designs have been proposed for segment-fabrication of the high-temperature superconducting (HTS) helical coils in the FFHR-d1 helical fusion reactor, joint-winding of the HTS coils wound by connecting conductor segments, and the ‘remountable’ HTS magnet (here ‘remountable’ means being able to mount and demount repeatedly) assembled from coil segments with demountable joints. A bridge-type mechanical lap joint and mechanical edge joint are planned to be applied to those two designs, respectively. This paper presents the progress in electrical and mechanical performances of the mechanical joints of HTS conductors and methods to evaluate their quality for segment-fabrication of HTS helical coils. R&D of joint performance has progressed during this decade and it shows acceptable performance for the HTS helical coils. The contact-probing current transfer length method and x-ray computer tomography scanning are promising for quality assessment of the joints and HTS tapes.

Novel features of the helical volumetric neutron source FFHR-b2

The compact helical fusion reactor FFHR-b2 is a multipurpose HElical Volumetric Neutron Source available as the component test facility to test the divertor and blanket systems. In the latest design, three new technologies of the high-temperature superconducting conductor, the ceramic pebble divertor, and the cartridge-type blanket are adopted. Tin-based functional liquid metals (FLMs) including lithium are considered as the working fluid for the blanket. The FLMs have multi functions of low vapor pressure, low density, low melting point, low corrosiveness, high tritium breeding ratio, and so on. The target construction cost of the FFHR-b2 has been set to 200 billion JPY. The main target of FFHR-b2 is to demonstrate the fusion gain larger than one. To achieve this in a relatively small device, the tangential neutral beam injection (NBI) heating with a moderate energy of 80 keV is adopted for the FFHR-b2. The beam-plasma fusion with 5 MW of NBI heating can produce a fusion power of 5 MW that is larger than the absorbed power of 3 MW.

Visualization of plasma shape in the LHD-type helical fusion reactor, FFHR, by a deep learning technique

Visualization of plasma shape in the LHD-type helical fusion reactor, FFHR, by a deep learning technique

Effect of the Pitch Modulation of Helical Coils on the Core Plasma Performance of the LHD-Type Helical Fusion Reactor

Effect of the Pitch Modulation of Helical Coils on the Core Plasma Performance of the LHD-Type Helical Fusion Reactor

Improvement of Superconducting Linear Acceleration System for Pellet Injection: Optimization of Current Profile

The improvement of the acceleration performance of a superconducting linear acceleration (SLA) system for injecting the pellet container into the helical fusion reactor has been investigated numerically. To this end, a numerical code used in the finite element method and the non-dominated sorting genetic algorithms-II has been developed for analyzing both the shielding current density in an HTS film and the dynamic motion of the film. In addition, the current profile of the acceleration coil is optimized by the normalized Gaussian network (NGnet) method. The results of the computations show that when the current profile of the outer coil is optimized by using the NGnet method, the pellet velocity is greatly faster than that for the homogeneous current profile. Moreover, the speedup ratio decreases monotonously as the coil length increases. If the length of the outer coil is more than twice that of the inner coil, the number of the filaments outside the inner coil reduces extremely. This result means that the amount of the filament and the energy consumption can be reduced compared with the homogeneous current profile.

Plasma irradiation experiment on the metal pebble flow in the TPDsheet-U

A new divertor concept REVOLVER-D2 that uses metal pebbles as a divertor target has been proposed for the LHD-type helical fusion reactor FFHR-c1. Demonstration of sufficient plasma shielding rate is one of the important issues to show the feasibility of the REVOLVER-D2. For this purpose, an experiment on the plasma irradiation to the metal pebble flow has been conducted in the sheet plasma device TPDsheet-U in Tokai University. Reduction of the ion flux and electron density in the metal pebble flow was observed and it was confirmed that the metal pebble flow can play the role of the divertor target. The measured shielding rate is consistent with the prediction by the model based on the Lambert-Beer-Bouger’s law. Achievement of the shielding rate of > 72 % is expected in the FFHR-c1 using the nozzle with the size of > 310 mm × 310 mm and the pebble flow with the diameter of 5 mm.

Topology optimization for superconducting magnet system in helical fusion reactor

In magnetic-confinement-type fusion reactors, the strong electromagnetic force generated by the superconducting coil is of the order of several tens of MN/m. To support this huge force, a strong and heavy coil support structure is required. The total weight of each magnet designed for conventional fusion reactors or experimental devices is several times greater than the ideal value. In recent years, structural optimization has advanced, owing to improvements in computer-aided design and finite-element analysis. For instance, the topology optimization method contributes novel designs that overturn common sense. In the present study, topology optimization is applied to the design of the magnet system used in a helical fusion reactor to reduce the weight of the magnet while maintaining the mechanical soundness of the component. A weight reduction of more than 25% from the conventional design could be achieved.

Conceptual design of a compact helical fusion reactor FFHR-c1 for the early demonstration of year-long electric power generation

Conceptual design of a compact large helical device (LHD)-type helical fusion reactor FFHR-c1 has been conducted. This design focuses on a year-long electric power generation with as small a reactor size as possible by adopting the operation with auxiliary heating and innovative ideas for the design of the superconducting magnet, divertor and blanket system. The primary design parameters, which have been selected by extrapolation from the previous design FFHR-d1B using simple scalings, has been examined by the systems code and the 1D integrated physics analysis code from the viewpoint of plant power balance and the achievable plasma operation regime. The design feasibility of the proposed design point has been confirmed with the physics parameters that are consistent with the LHD experiment. Further improvement of the plasma performance is expected by the optimisation of the helical coil winding law. Although intensive R&D is needed to realise the innovative ideas of the engineering design, this design ensures the path to helical commercial power plants and provides more options and increases the probability of solving critical issues of fusion reactors.

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.

The strategy toward realization of the helical fusion reactor FFHR-c1

Design studies on the helical fusion reactor FFHR-c1 have progressed. The main goal of the FFHR-c1 is to demonstrate one-year steady-state sustainment of the fusion plasma with self-produced fuel and electricity. The major radius of the plasma, R, is ˜10 m and the magnetic field strength at the plasma center, B, is ˜8 T. The magnet coil arrangement is basically similar to that of the LHD. The fusion gain during the one-year operation will be ˜15 with the fusion output of ˜370 MW. Three new technologies of the high-temperature superconducting magnet coils, the cartridge-type liquid blanket, and the fusible metal pebble divertor are adopted. Including these three, we have defined 22 important issues. The strategy to efficiently address the 22 issues has been also discussed. As a result, we propose a step-by-step approach by developing the FFHR-a1 (R ˜ 2.5 m and B ˜ 4 T) and the FFHR-b1 (R ˜ 3.6 m and B ˜ 6 T), before the FFHR-c1. The FFHR-b1 is a HElical Volumetric Neutron Source (HEVNS) driven by the neutral beam injection heating, and the FFHR-a1 corresponds to the pilot HEVNS.

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.

Advanced Helical Plasma Research towards a Steady-State Fusion Reactor by Deuterium Experiments in Large Helical Device

The Large Helical Device (LHD) is one of the world’s largest superconducting helical system fusion-experiment devices. Since the start of experiments in 1998, it has expanded its parameter regime. It has also demonstrated world-leading steady-state operation. Based on this progress, the LHD has moved on to the advanced research phase, that is, deuterium experiment, which started in March 2017. During the first deuterium experiment campaign, an ion temperature of 10 keV was achieved. This was a milestone in helical systems research: demonstrating one of the conditions for fusion. All of this progress and increased understanding have provided the basis for designing an LHD-type steady-state helical fusion reactor. Moreover, LHD plasmas have been utilized not only for fusion research, but also for diagnostics development and applications in wide-ranging plasma research. A few examples of such contributions of LHD plasmas (spectroscopic study and the development of a new type of interferometer) are introduced in this paper.

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.

The Large Helical Device: Entering Deuterium Experiment Phase Toward Steady-State Helical Fusion Reactor Based on Achievements in Hydrogen Experiment Phase

The large helical device (LHD) is one of the world’s largest superconducting helical system fusion-experiment devices. Since the start of experiments in 1998, LHD has extended its parameter regime, aiming at achievement of the reactor-relevant plasma conditions and the exploration of related plasma physics in helical-type magnetic configurations. The LHD has also demonstrated its inherent advantage for steady-state operation. Based on these leading developments of helical plasma research, LHD has progressed to the advanced research phase, that is, the deuterium experiment that started in March 2017. It is expected that plasma parameters should be extended toward more reactor-relevant regime, and the related physics research is allowed in such extended regime. Taking this opportunity, parameter extensions such as density, temperature, and steady-state operation achieved in the hydrogen experiment phase are overviewed, along with the initial highlighted results in the very first deuterium experiment campaign in 2017. The design activity of LHD-type steady-state helical fusion reactor is also briefly introduced.

Maintainability of the helical reactor FFHR-c1 equipped with the liquid metal divertor and cartridge-type blankets

The maintenance scheme in the compact helical fusion reactor, FFHR-c1, equipped with the liquid metal divertor, REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) and the cartridge-type molten salt blankets, CARDISTRY-B (CARtridges Divided and InSerTed RadiallY − Blanket), is investigated. The magnetic configuration of the compact variant FFHR-c1 is similar to that of the Large Helical Device (LHD), while the device size is 2.8 times enlarged from LHD and a strong magnetic field strength of ∼8 T at the plasma center is adopted. The maintenance of the REVOLVER-D is simpler than that of the helical divertor with a complicated structure as seen in the LHD. In the REVOLVER-D, showers of molten tin are injected into the ergodic layer at 10 inner ports. To circulate the molten tin, 10 sets of the shower system including a liquid metal pump, ducts, a showerhead, a pool, and a heat exchanger, are installed. These can be replaced with simple up/down motions. The CARDISTRY-B consists of 320 tritium breeding blanket cartridges, which are toroidally segmented every two degrees. These cartridges are maintained by using heavy manipulators with a simple 4DOF motion at a constant toroidal angle.

Prospect Toward Steady-State Helical Fusion Reactor Based on Progress of LHD Project Entering the Deuterium Experiment Phase

Large Helical Device (LHD) is one of the world largest superconducting fusion experiment devices, having demonstrated its inherent advantage for steady-state operation since the start of experiments in 1998. LHD has also demonstrated reliable operation of the large-scale superconducting magnet system for almost two decades. Development of the challenging heating systems, such as negative-ion-based neutral beam injection (NBI), high-power and high-frequency electron cyclotron heating, and steady-state ion cyclotron heating, have led to wide-ranging physics and engineering achievements. LHD has progressed to the next stage, that is, the deuterium experiment starting in March 2017, which should further extend plasma parameters toward reactor-relevant regime. For establishing firm basis for designing steady-state helical fusion reactor, advanced physics research, such as on isotope effect, energetic particle confinement, and plasma–wall interaction, will be intensively performed in the deuterium experiments. In an engineering aspect, the upgrade of NBI system has been carried out in preparation to the deuterium experiment, and it should contribute to future NBI development for fusion reactors including ITER. For enhancement of the particle control, the closed divertor system has been installed with pumping capability. Diagnostics for neutron measurements are newly developed and installed for the deuterium experiment. Aligned with all the progress of LHD project in terms of engineering and physics aspects, the conceptual design activity of the LHD-type helical fusion reactor, FFHR-d1, has been programmatically conducted. In parallel to the design study, engineering research and development for the component development have been performed, including those based on employing challenging ideas such as high-temperature superconductor, liquid metal ergodic divertor, and molten-salt breeder blanket. The present status of LHD project entering the deuterium experiment phase is overviewed with putting emphasis on the engineering aspects, and then the engineering research and development activities toward steady-state helical fusion reactor are described.

Thermal and flow analyses on the cartridge-type blanket CARDISTRY-B for the helical fusion reactor FFHR-c1

Thermal and flow analyses on the cartridge-type molten salt blanket named the CARDISTRY-B (CARtridges Divided and InSerTed RadiallY – Blanket) for the compact helical fusion reactor FFHR-c1 have been carried out. As a result, it is shown that the maximum local temperature in the cartridges can be kept below 550 °C, which is an allowable temperature of the structure material, under the neutron wall loading condition of FFHR-c1. The CARDISTRY-B is composed of the tritium breeder (Breeding Blanket: BB) and the neutron shield (Shielding Blanket: SB). Both of them are toroidally segmented every two degrees. From a safety perspective, the molten salt FLiNaBe (melting point ∼305 °C) is chosen as the first option of the breeder material. In the case of FLiNaBe BB cartridge made of RAFM (Reduced Activation Ferritic/Martensitic steel), it is necessary to suppress the maximum temperature of BB cartridges to 550 °C or below. In this study, a finite volume method code considering beryllium (Be) pebbles inside a flow channel has been used to evaluate the cooling characteristics of a BB cartridge. By estimating the pressure drop through pebbles with the porous medium model, we have investigated how the temperature of BB cartridge change depends on the mass flow rate of FLiNaBe. The volumetric heating profile was modeled according to the neutronics calculation results obtained by the MCNP (Monte Carlo Neutron and Photon Transport) code.