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Although ICRF heating has achieved the high heating efficiency necessary to achieve high-performance plasmas, it has not overcome the reliability and economic problems associated with the antenna structure inside the vacuum vessel in fusion reactors. We suggested a slit ICRF antenna that uses the blanket surface as a transmission line to solve these problems. With a single slit ICRF antenna with a width of 3 m and a height of 15 cm, the electric field strength to the magnetic field direction was successfully suppressed to 5 kV/cm when 20 MW of power radiation was achieved from the single slit. The slit ICRF antenna had a bending angle in the electromagnetic wave transmission path to prevent direct neutron impact on the first wall and a vacuum gate from rapidly preventing water or air leakage accidents. The slit ICRF antenna has a simple structure that allows heating at high power density while minimizing blanket volume reduction.

In the Large Helical Device (LHD), the development of high-power and long-pulse Ion Cyclotron Range of Frequencies (ICRF) heating system is ongoing. The developed Field-Aligned-Impedance-Transforming (FAIT) antenna has the potential for high-power injection of more than 1.8 MW. Here, to achieve this injection power, a power combination system was developed. An optimized power combiner was designed by repeated simulations, and then was fabricated and installed in the ICRF transmission system. Control of the power and the phase of incident waves into the input ports of the power combiner is important for the power combination. Therefore, a real-time control system was developed, and prompt reduction of power loss was demonstrated. As a result, combined powers of more than 2 MW for 6 s and 1 MW for 10 min were successfully achieved.

The transmission line is one of the most important components of ion cyclotron range of frequencies (ICRF) heating devices. In the case of unexpected trouble on the line, such as a breakdown, immediate power-off is necessary in order to avoid severe damage on the line. Breakdowns are difficult to detect with a reflection monitor, since the reflection may originate from a change in the antenna-plasma coupling. In the Large Helical Device (LHD), a Fault Detection System (FDS) for the transmission line was developed, which detects the breakdown utilizing the unbalance of three signals from the both ends of the line. For the precise balancing in the normal condition, the calibration is iteratively conducted. FDS is insensitive to the change of the antenna impedance, therefore, FDS can detect breakdown clearly. Frequency shift is also detectable with the FDS applied to a long transmission line. Therefore, the self-oscillation accompanying frequency shift could be detected in addition to breakdown.

For the series production of the divertor outer vertical target (OVT) of the ITER, approximately 200,000 tungsten (W) monoblocks as plasma-facing material are planned to be manufactured. However, thus far, the variation in the properties of the W monoblocks has not been investigated. In this study, the chemical impurity level, material properties, and response to cyclic heat loading (i.e., durability and surface deformation) of W monoblocks manufactured using W plates supplied by A.L.M.T. Corp., Japan (ALMT) and Nidec Sankyo CMI Corp., Japan (NSCM) were investigated. On examining the manufactured lots (four from ALMT and two from NSCM), it was concluded that the variation in the material properties might have been caused by variations in the impurity level of the raw powder, contamination during manufacturing, and changes in the processing parameters. The results of the high heat flux test indicated that all the examined lots met the acceptance criteria. Hence, the variations observed in the chemical impurity level and material properties (e.g., density, Vickers hardness, and grain size) of the W plates supplied by ALMT and NSCM were acceptable. Therefore, W from the two suppliers will be employed in the full-scale plasma-facing unit prototype, which will be used for the final qualification of the material and OVT manufacturing technology.

ITER divertor outer vertical target (OVT) consists of 22 plasma facing units (PFUs) and 2 stainless support structures (SSSs) including a complex internal structure for coolant water. Pressure drop and flow rate in each PFU are identified as important parameters for acceptable range in terms of the pumping power of cooling water system and performance of heat removal. In this study, computational fluid dynamics (CFD) and a transient reactor analysis code (TRAC-PF1) which is simpler simulation than the CFD have been performed to estimate the pressure drop and flow distribution in 11 PFUs mounted on an OVT SSS (half of the OVT). Results of both prediction methods showed the proper pressure drop and the homogeneous flow distribution within acceptable range of design requirement. To verify these numerical simulations, the results were compared to experimentally measured values in the flow test. This study demonstrated that the TRAC-PF1 as well as the CFD are applicable tools for further optimization of the divertor OVT cooling structure in view of manufacturing.

ITER divertor outer vertical target (OVT) consists of 22 plasma facing units (PFUs) and 2 stainless support structures (SSSs) including a complex internal structure for coolant water. Pressure drop and flow rate in each PFU are identified as important parameters for acceptable range in terms of the pumping power of cooling water system and performance of heat removal. In this study, computational fluid dynamics (CFD) and a transient reactor analysis code (TRAC-PF1) which is simpler simulation than the CFD have been performed to estimate the pressure drop and flow distribution in 11 PFUs mounted on an OVT SSS (half of the OVT). Results of both prediction methods showed the proper pressure drop and the homogeneous flow distribution within acceptable range of design requirement. To verify these numerical simulations, the results were compared to experimentally measured values in the flow test. This study demonstrated that the TRAC-PF1 as well as the CFD are applicable tools for further optimization of the divertor OVT cooling structure in view of manufacturing.

The refined design of full-W ITER divertor outer vertical targets (OVT) has the combination of plasma-facing units (PFUs) variants with bevelled W monoblocks to prevent leading edge overheating. Full-length PFU prototypes have withstood against the repetitive heat load of 20MW/m2 and they met the requirements for W technologies qualification programed by the IO. Upon the successful results, JADA has launched series of qualification programs toward the mass production of components. Furthermore, JADA successfully verified the water flow design inside the OVT is fulfilling the ITER requirements. The flow test was done using a OVT prototype with dummy cooling tubes.

Newly developed ion cyclotron range of frequencies (ICRF) antennas were fabricated and installed in the Large Helical Device (LHD). The antennas, called field-aligned-impedance-transforming (FAIT) antennas, were designed based on former handshake form (HAS) antennas for high power and steady-state operation. The FAIT antennas demonstrated higher loading resistance and lower temperature on ceramic feed-through than HAS antennas. Antenna impedance and injected power were measured precisely using a directional coupler installed between the impedance matching device and antenna. It was shown that an injection power of 1.8 MW or a power density of 15 MW/m2 would be possible for short-pulse discharges. The mutual coupling between two FAIT antennas is small enough, since any current phase is possible without adjusting the impedance matching device. Although the FAIT antennas are aligned in a poloidal direction, a high heating efficiency of 85% was achieved with the proper minority concentration ratio.

For steady state operation, the feedback control of plasma density and heating power system was developed in the Large Helical Device (LHD). In order to achieve a record of the long pulse discharge, stable plasma density and heating power are needed. This system contains the radio frequency (RF) heating power control, interlocks, gas fueling, automatic RF phase control, ion cyclotron range of frequency (ICRF) antenna position control, and graphical user interface (GUI). Using the density control system, the electron density was controlled to the target density and using the RF heating power control system, the RF power injection could be stable. As a result of using this system, we achieved the long pulse up to 48 min in the electron density of more than 1 × 10 19 m −3 . Further, the ICRF hardware experienced no critical accidents during the 17th LHD experiment campaign in 2013.