Design Window Analysis for the Helical DEMO Reactor FFHR-d1

Conceptual design activity for the LHD-type helical DEMO reactor FFHR-d1 has been conducted at the National Institute for Fusion Science under the Fusion Engineering Research Project since FY2010. In the first step of the conceptual design process, design window analysis was conducted using the system design code HELIOSCOPE by the “Design Integration Task Group”. On the basis of a parametric scan with the core plasma design based on the DPE (Direct Profile Extrapolation) method, a design point having a major radius of 15.6 m and averaged magnetic field strength at the helical coil winding center of 4.7 T was selected as a candidate. The validity of the design was confirmed through the analysis by the related task groups (in-vessel component, blanket, and superconducting magnet)

Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, \u27basic\u27 and \u27challenging.\u27 Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor

Fuel Particle Balance Study in FFHR DEMO Reactor

Tritium particle balance in the FFHR DEMO reactor is investigated with consideration of the fueling efficiency by pellet injection system, retention loss in a vacuum vessel and permeation loss from the fuel processing system. In order to satisfy the fuel balance and the tritium safety management, it was necessary to suppress the tritium retention rate to be 10?5 and the DFs in the tritium cycle systems to above 107 with the tritium breeding ratio of 1.08. The processing throughput for the tritium processing system is estimated to be about 400 mol/h, which is almost same as the throughput of the fuel stream for the ITER. Therefore, the tritium processing system for vacuum exhaust gas for the DEMO will not be necessary to improve the system for the ITER further. On the other hands, the significant development of the tritium processing system for the effluent disposal and the waste materials from the safety aspect and the social acceptance will be required toward the DEMO reactor

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

The effect of the pitch modulation of the helical coils on the core plasma performance of the LHD-type helical fusion reactor has been examined. The analysis of the MHD stability and neoclassical transport for the pitch modulation α = 0.0 and 0.1 has been conducted based on the finite-beta equilibrium calculated by the HINT code. It was found that the MHD stability is clearly improved without deteriorating the energy transport property by changing the pitch modulation α from 0.1 to 0.0. The reachable operation region expands to the higher density and the expected fusion gain can increase from ∼10 to ∼20. Because the change of the pitch modulation α from 0.1 to 0.0 requires only a slight change in the shape of the helical coils, the engineering design including the maintenance method that has been examined for the reactor with α = 0.1 can be applied without a major modification

Neutronics Investigations for Helical DEMO Reactor FFHR-d1

Radiation shielding and tritium breeding performances of the helical DEMO reactor FFHR-d1 have been investigated for the present proposed reactor component configuration. Since the core plasma position shifts to inboard side of the torus, the total thickness of the inboard breeding blanket and radiation shield of FFHR-d1 is limited to ?70 cm. To simulate the geometric features of the helical reactor, three-dimensional neutronics calculation model consisted of ?4,000 cells have been prepared for the neutron and gamma-ray transport calculations using the MCNP code. It has been confirmed that the present radiation shield configuration with WC (tungsten carbide) and FS (ferritic steel) + B4C layers would provide sufficient shielding performance for the helical coils. The tritium breeding ratio (TBR) of 1.08 has been obtained with a Flibe+Be/FS breeding blanket for the present component configuration of FFHR-d1