A. Sagara

Concept of magnet systems for LHD-type reactor

Heliotron reactors have attractive features for fusion power plants such as having no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered to be the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, a major radius of plasma of 14?17?m with a central toroidal field of 6?4?T is needed to attain the self-ignition condition with a blanket space thicker than 1.1?m. The stored magnetic energy is estimated at 120?140?GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress are comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than the 1000?MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is more than 150?m, that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with a small extension of the ITER technology.

Design and development of high-temperature superconducting magnet system with joint-winding for the helical fusion reactor

An innovative winding method is developed by connecting high-temperature superconducting (HTS) conductors to enable efficient construction of a magnet system for the helical fusion reactor FFHR-d1. A large-current capacity HTS conductor, referred to as STARS, is being developed by the incorporation of several innovative ideas, such as the simple stacking of state-of-the-art yttrium barium copper oxide tapes embedded in a copper jacket, surrounded by electrical insulation inside a conductor, and an outer stainless-steel jacket cooled by helium gas. A prototype conductor sample was fabricated and reached a current of 100 kA at a bias magnetic field of 5.3 T with the temperature at 20 K. At 4.2 K, the maximum current reached was 120 kA, and a current of 100 kA was successfully sustained for 1 h. A low-resistance bridge-type mechanical lap joint was developed and a joint resistance of 2 nΩ was experimentally confirmed for the conductor sample.

Progress of the Design of HTS Magnet Option and R&D Activities for the Helical Fusion Reactor

The high-temperature superconducting magnet option is being explored in the conceptual design studies of the LHD-type helical fusion reactor FFHR-d1. A 100 kA-class conductor is being developed by simply stacking REBCO tapes in a copper and stainless-steel jacket. One of the design options of the HTS conductor includes internal insulation so that the windings do not require vacuum pressure impregnation process. Innovative winding method of the huge helical coils is being investigated based on the segment fabrication of half-helical-pitch conductors by developing a bridge-type mechanical lap joint. A “30 kA-class” prototype conductor sample was fabricated using GdBCO tapes and successfully tested. The critical current was measured at various temperatures at 4.2-40 K and magnetic field <; 8 T. The joint resistance was evaluated by changing the applied stress. These experimental results are boosting the HTS magnet design of FFHR-d1.

Measurement and Analysis of Critical Current of 100-kA Class Simply-Stacked HTS Conductors

Based on the successful plasma experiments in the Large Helical Device (LHD), design activities of the LHD-type helical fusion reactor FFHR-d1 are progressing at National Institute for Fusion Science (NIFS). A 100 kA current capacity is required for the winding conductor under the maximum magnetic field of ~12 T. The high-temperature superconductor (HTS) is a promising option for the helical coil conductor. For the development of such a HTS conductor suitable for the helical fusion reactor, we fabricated 30 kA-class HTS conductor samples, and the excitation tests were successfully carried out. We then fabricated and tested a 100-kA class HTS conductor. The conductor sample is a one-turn short-circuit coil with a race-track shape having a bridge-type mechanical lap joint. The transport current of the sample was induced by changing the external magnetic field, then the critical current of the sample was measured. A numerical analysis of the critical current is being performed by self-consistently solving the spatial distributions of the current density and magnetic field among the simply-stacked HTS tapes to verify the measured critical current of the samples. The critical current characteristics of a single HTS tape is evaluated by the percolation model in the precise analysis.

Critical Current Measurement of 30 kA-Class HTS Conductor Samples

Design activities on the helical-type fusion DEMO reactor, FFHR-d1, are progressing at NIFS. A 100 kA current-capacity is required for the helical coil conductors under the maximum magnetic field of ~ 13 T. High-temperature superconducting conductor has been proposed as one of the conductor options for the FFHR-d1 magnet. In this study, a 30 kA class HTS conductor sample has been fabricated and tested. The sample had no current feeders and the current was induced by changing the background magnetic field generated by the 9 T split coils in the cryostat. Rogowski coils and Hall probes were used for the measurement of the transport current of the sample. The critical current of the sample was measured at various temperatures and bias magnetic fields. To verify the self-field effect of the sample, a numerical analysis was performed by considering the current and magnetic field distribution among the tapes self-consistently. The analysis result was compared with the experimental observation.

An evaluation of fusion gain in the compact helical fusion reactor FFHR-c1

A new procedure to predict achievable fusion gain in a sub-ignition fusion reactor is proposed. This procedure uses the direct profile extrapolation (DPE) method based on the gyro-Bohm model. The DPE method has been developed to predict the radial profiles in a fusion reactor sustained without auxiliary heating (i.e., in the self-ignition state) from the experimental data. To evaluate the fusion gain in a fusion reactor sustained with auxiliary heating (i.e., in the sub-ignition state), the DPE method is modified to include the influence of the auxiliary heating. The beta scale factor from experiment to reactor is assumed to be 1. Under this assumption, it becomes reasonable to apply the magnetohydrodynamic (MHD) equilibrium (which is calculated to reproduce the experimental data) to the reactor. At the same time, the MHD stability of the reactor plasma is also guaranteed to a certain extent since that beta was already proven in the experiment. The fusion gain in the helical type nuclear test machine FFHR-c1 has been evaluated using this modified DPE method. FFHR-c1 is basically a large duplication of the Large Helical Device (LHD) with a scale factor of 10/3, which corresponds to the major radius of the helical coils of 13.0 m and the plasma volume of ~1000 m3. Two options with different magnetic field strengths are considered. The fusion gain in FFHR-c1 extrapolated from a set of radial profile data obtained in LHD ranges from 1 to 7, depending on the profiles used together with the assumptions of the magnetic field strength and the alpha heating efficiency.

Numerical analysis of MHD flow in remountable first wall

Abstract New concept of remountable first wall using liquid flow has been proposed as a different approach of liquid wall concepts like APEX (Abdou, the APEX team, Fusion Eng. Des. 54 (2001) 181). In this study, the numerical analysis of MHD flow has been carried out to evaluate effect of coating on the MHD pressure drop. In the proposed concept the liquid layer is sandwiched by two metal plates with two side ribs to compose the flow channel. The numerical results indicate that coating the whole rib surface has good performance as three surface coating to reduce the MHD pressure drop. In both cases of flibe and lithium coolants, the solid metal (HT-9) wall can be used by coating with the insulator whose electric conductivity ratio ( σ coating / σ HT-9 ) is less than 10 −6 under 6 T.

Feasibility Study for Flibe TBM Based on Thermofluid Analysis

Abstract By changing the composition ratio in Flibe to decrease its melting temperature, it becomes possible to design the TBM under the temperature design limits for ferritic steel. The accompanied demerit due to the increase in viscosity and degradation in heat transfer performance is overcome by introducing sphere-packed pipe as the first wall. The empirical correlation for heat transfer performance is derived for several sizes and materials of the spheres. Through the present analysis, there exist design windows for the Flibe TBM. This possibility is strongly linked to the demo reactor development since the structural material development for higher temperature condition can lead to the usage of Flibe with higher melting temperature and better heat transfer performance, which could be available under higher heat flux in the demo reactor.

Cost Assessment of Fusion Reactor with External Conductor Systems

Abstract Fusion reactors with external SC (superconducting) magnet systems have the possibility of high beta-value without plasma current for plasma confinement. These reactors, however, have complex configurations and systems, which leads to reduction of these plant availabilities together with unstable maintainability. Through this research the influence on the complex system to its availability is evaluated to demonstrate the effect of using remountable systems on COE.