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A new cartridge-type blanket named the CARDISTRY-B is proposed for the helical fusion reactor FFHR-d1. This blanket is composed of the neutron shield and the tritium breeder using molten salt. Both of these are toroidally segmented every two degrees. At each toroidal angle, the segmented parts are divided further into several cartridges in order to make it possible to assemble these cartridges after completion of the superconducting magnet coils. The neutron shield is basically assembled by using mortise and tenon prepared on each of the cartridges and the lower port, instead of the wide area welding. After assembly, the plasma side of the neutron shield is welded to form a vacuum vessel. Another side of the neutron shield facing on the superconducting magnet coils is covered with the thermal shield, which was already attached before assembly. The tritium breeder cartridges can be replaced without cutting or welding of cooling pipes inside the vacuum vessel, where severe radiation dose is expected. Details of the CARDISTRY-B, including the results of motion analysis for all cartridges and estimation of the cartridge weight, are discussed., source:https://doi.org/10.1585/pfr.12.1405017, identifier:0000-0002-7901-6077

The required pumping power for the liquid metal ergodic limiter/divertor REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor), which is considered as an optional divertor design for the LHD-type helical fusion reactor FFHR-d1, has been estimated with a consideration of the change in the strength of transverse magnetic field along the flow channel. It is found that the pressure drop and the required pumping power can be suppressed to an acceptable level by using insulated ducts., source:https://doi.org/10.1585/pfr.12.1405016, identifier:0000-0001-6028-8980

The cartridge-type helical blanket system called the CARDISTRY-B has been proposed for the helical fusion reactor FFHR-d1. The CARDISTRY-B is aimed at easy construction and maintenance. However, there remain many issues in the design. For example, these include a large number of cartridges, no tangential ports, low compatibility with the helical divertor, and others. To solve these issues, an improved design of the cartridge-type blanket is proposed and named the CARDISTRY-B2. This is composed of the tritium breeding blanket (BB) filled with the flowing molten salt and the neutron shielding blanket (SB) filled with the tungsten-carbide. The number of the BB (or, SB) cartridges in the 1/10 section of the full torus are reduced from 32 to 17 (or, from 44 to 17). Tangential ports for the neutral beam injection heating are newly equipped together with the multi-purpose blanket cartridge. Compatibility with the helical divertor is also improved. The BB cartridges are fixed on the fixing units set on the top of the SB. No other part of the BB is touching the SB except the fixing ribs. The SB cartridges surround the superconducting magnet coils and work as the cryostat inside the vacuum vessel. This makes the large bellows for the large maintenance ports unnecessary. Details of the CARDISTRY-B2 designed for a small helical fusion reactor FFHR-b1 are described in this paper., source:https://doi.org/10.1585/pfr.14.1405163, identifier:0000-0002-7901-6077

Tearing mode is studied in a heliotron plasma by varying the net toroidal current. Ideal internal kink mode can be unstable when two resonant magnetic surfaces are generated. The tearing mode is found at the current below the marginal stability value for the ideal internal kink mode. The mode structure is localized around the inner resonant surface like the internal kink mode. The growth rate and the mode structure can be affected by the unstable internal kink mode. Tearing and resistive interchange hybrid mode is also obtained at a finite beta., identifier:0000-0002-7698-0223

The effects of the deformation of the flux surfaces into a “D shape” on vacuum heliotron configurations are examined numerically. In this study, one poloidal coil on the mid plane just inside the helical coils is assumed to be installed so that the current in the coil can deform the surface shape. Such deformation gives the advantage that the clearance between the helical coil and the outermost surface can be enlarged, which is crucial for the divertor action in heliotrons. The effects of the deformation on other vacuum properties are also discussed, such as rotational transform, magnetic well, and |B|min contour. We also focus on such effects on downsized configurations., source:https://doi.org/10.1585/pfr.14.3403100, identifier:0000-0002-7698-0223

Effects of poloidal shear flow on the stability of interchange modes in a Large Helical Device (LHD) configuration are numerically studied. Three-dimensional (3D) numerical codes are utilized for the equilibrium and stability calculations. A static equilibrium is employed and a model poloidal flow as a flux function is incorporated in the initial perturbation. The results show that the initially applied flow can suppress the growth of the interchange mode if the flow is sufficiently large., source:https://doi.org/10.1585/pfr.11.2403035, identifier:0000-0002-7698-0223

The rotation of resistive interchange instabilities excited in the periphery of plasmas compared to plasma flows under various plasma conditions in the Large Helical Device. The observed magnetohydrodynamic modes exhibit nominal rotation in the electron diamagnetic direction in the laboratory frame. We assumed that the ion (electron) flows consist of the E × B and ion (electron) diamagnetic drift flows. These drift flows were individually evaluated and geometrically compensated using a finite beta equilibrium. Experiments with different neutral beam injection conditions and different electron densities show that the frequencies of the observed modes were quantitatively consistent with those of the electron flow., source:https://doi.org/10.1585/pfr.8.1402123, identifier:0000-0003-3754-897X

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., source:T. Goto et al 2019 Nucl. Fusion 59 076030, source:https://doi.org/10.1088/1741-4326/ab15c3, identifier:0000-0001-6028-8980

Integrated physics analysis of plasma operation scenario of the compact helical reactor FFHR-c1 has been conducted. The DPE method, which predicts radial profiles in a reactor by direct extrapolation from the reference experimental data, has been extended to implement the equipartition effect. Close investigation of the plasma operation regime has been conducted and a candidate plasma operation point of FFHR-c1 has been identified within the parameter regime that has already been confirmed in LHD experiment in view of MHD equilibrium, MHD stability and neoclassical transport., source:T Goto et al 2018 Plasma Phys. Control. Fusion 60 074001, source:https://doi.org/10.1088/1361-6587/aabd51, identifier:0000-0001-6028-8980

This study closely investigates the plasma operation scenario for the LHD-type helical reactor FFHR-d1 in view of MHD equilibrium/stability, neoclassical transport, alpha energy loss and impurity effect. In 1D calculation code that reproduces the typical pellet discharges in LHD experiments, we identify a self-consistent solution of the plasma operation scenario which achieves steady-state sustainment of the burning plasma with a fusion gain of Q ~ 10 was found within the operation regime that has been already confirmed in LHD experiment. The developed calculation tool enables systematic analysis of the operation regime in real time., source:T. Goto et al 2017 Nucl. Fusion 57 066011, source:https://doi.org/10.1088/1741-4326/aa6870, identifier:0000-0001-6028-8980

1D physics analysis of the plasma start-up scenario of the large helical device (LHD)-type helical reactor FFHR-d1 was conducted. The time evolution of the plasma profile is calculated using a simple model based on the LHD experimental observations. A detailed assessment of the magnetohydrodynamic equilibrium and neo-classical energy loss was conducted using the integrated transport analysis code TASK3D. The robust controllability of the fusion power was confirmed by feedback control of the pellet fuelling and a simple staged variation of the external heating power with a small number of simple diagnostics (line-averaged electron density, edge electron density and fusion power). A baseline operation control scenario (plasma start-up and steady-state sustainment) of the FFHR-d1 reactor for both self-ignition and sub-ignition operation modes was demonstrated., source:T. Goto et al 2015 Nucl. Fusion 55 063040, source:https://doi.org/10.1088/0029-5515/55/6/063040, identifier:0000-0001-6028-8980

The mechanism of the partial collapse observed in the experiment with the background magnetic field changing in the Large Helical Device (LHD) is numerically investigated with a nonlinear magnetohydrodynamics (MHD) simulation. Since the different timescales of the perturbations and the background field changing have to be treated simultaneously for the analysisof this plasma, a multi-scale simulation scheme is developed. The effect of the perturbation dynamics on the equilibrium pressure and rotational transform is taken into account in this scheme. The result indicates that the collapse is caused by the destabilization of an infernal-like mode due to the magnetic hill enhanced by the change of the background field. The mechanismof the reduction of the central beta observed after the partial collapse in the experiment is also analysed in relation to the effect of the background field changing., source:Ishiguchi K, Sakakibara S, Ohdachi S and Carreras B.A. 2015 Multi-scale MHD analysis of LHD plasma with background field changing Nuclear Fusion 55 043019., source:https://doi.org/10.1088/0029-5515/55/4/043019, identifier:0000-0002-7698-0223

A new liquid metal divertor system named the REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) is proposed for the helical fusion reactor FFHR-d1. The REVOLVER-D is composed of molten tin shower jets stabilized by internal flow resistances of wire/tape/chain. These shower jets are inserted into the ergodic layer surrounding the main plasma. Tin is selected as the liquid metal because of its low melting point, low vapor pressure, low material cost, and high safety. The liquid metal pumps, cryopumps, and turbo molecular pumps are installed in the central vacuum vessel connected to the main vacuum vessel via 10 inner ports equipped with maze neutron shields. Central solenoid coils made of high-temperature superconductors are installed inside the central vacuum vessel to shield the pumps from the strong magnetic field. The REVOLVER-D has a good possibility to satisfy important characteristics required for the divertor system in a fusion reactor, that is, high heat load tolerance, high maintainability, sufficient vacuum pump speed, high level of safety, and a small amount of radioactive wastes., source:https://doi.org/10.1016/j.fusengdes.2017.07.003, identifier:0000-0002-7901-6077

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., source:Junichi Miyazawa, Takuya Goto, Hitoshi Tamura, Teruya Tanaka, Nagato Yanagi, Tananori Murase, Ryuichi Sakamoto, Suguru Masuzaki, Takeru Ohgo, Akio Sagara, Maintainability of the helical reactor FFHR-c1 equipped with the liquid metal divertor and cartridge-type blankets, Fusion Engineering and Design, Volume 136, Part B, 2018, Pages 1278-1285, source:https://doi.org/10.1016/j.fusengdes.2018.04.118, identifier:0000-0002-7901-6077

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., source:Yanagi, N., Goto, T., Miyazawa, J. et al. Progress in the Conceptual Design of the Helical Fusion Reactor FFHR-d1. J Fusion Energ 38, 147–161 (2019). https://doi.org/10.1007/s10894-018-0193-y, source:https://doi.org/10.1007/s10894-018-0193-y, identifier:0000-0002-4489-8241

A simultaneous high ion temperature (Ti) and high electron temperature (Te) regime was successfully extended due to an optimized heating scenario in the LHD. Such high-temperature plasmas were realized by the simultaneous formation of an electron internal transport barrier (ITB) and an ion ITB by the combination of high power NBI and ECRH. Although the ion thermal confinement was degraded in the plasma core with an increase of Te/Ti by the on-axis ECRH, it was found that the ion thermal confinement was improved at the plasma edge. The normalized ion thermal diffusivity ${{\chi}_{\text{i}}}/T_{\text{i}}^{1.5}$ at the plasma edge was reduced by 70%. The improvement of the ion thermal confinement at the edge led to an increase in Ti in the entire plasma region, even though the core transport was degraded., source:H. Takahashi et al 2017 Nucl. Fusion 57 086029, source:https://doi.org/10.1088/1741-4326/aa754b, identifier:0000-0001-6984-9174

The deuterium experiment started from March 2017 on the large helical device (LHD) as a part of the LHD high-performance upgrade project. The objectives of the deuterium experiment are: 1) to realize the high-performance operation; 2) to explore the physics of isotope effect; 3) to demonstrate the confinement capability of energetic particles in helical devices; and 4) to explore the research on plasma-material interactions using the benefit of stable steady-state operation ability of LHD. As preparations for deuterium experiment, the positive-ion-based neutral beam injectors (NBIs) are upgraded to increase their injection power in deuterium operation. On the other hand, the injection power of negative-ion-based NBI is deteriorated due to the isotope effect of negative-ion source. The neutron diagnostic and the exhaust detritiation system are newly installed for the deuterium experiment. The commissioning of LHD for the deuterium experiment is quite successful. The high ion temperature operation region is extended by the deuterium experiment. Ion temperature exceeding 9 keV is achieved with the deuterium operation of LHD., source:M. Osakabe et al., "Preparation and Commissioning for the LHD Deuterium Experiment," in IEEE Transactions on Plasma Science, vol. 46, no. 6, pp. 2324-2331, June 2018, doi: 10.1109/TPS.2018.2825423., source:https://doi.org/10.1109/TPS.2018.2825423, identifier:0000-0001-5220-947X