Achievement of reactor relevant plasma condition in Helical type magnetic devices and exploration in its related plasma physics and fusion engineering are the aim of the Large Helical Device (LHD) project. In the recent experiments on LHD, we have achieved ion-temperature of 8.1 keV at 1 × 1019 m−3 by the optimization of wall conditioning using long pulse discharge by Ion Cyclotron Heating (ICH). The electron temperature of 10 keV at 1.6 × 1019 m−3 was also achieved by the optimization of Electron Cyclotron Heating (ECH). For further improvement in plasma performance, the upgrade of the Large Helical Device (LHD), including the deuterium experiment, is planned. In this paper, the recent achievements on LHD and the upgrade of LHD are described.

Current Status of Large Helical Device and Its Prospect for Deuterium Experiment

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.

Preparation and Commissioning for the LHD Deuterium Experiment

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.

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

Design and commissioning of the exhaust detritiation system for the Large Helical Device

Abstract In recent deuterium experiments on the large helical device (LHD), we succeeded in expanding the temperature domain to higher regions for both electron and ion temperatures. Suppression of the energetic particle driven resistive interchange mode (EIC) by a moderate electron temperature increase is a key technique to extend the high temperature domain of LHD plasmas. We found a clear isotope effect in the formation of the internal transport barrier in high temperature plasmas. A new technique to measure the hydrogen isotope fraction was developed in the LHD in order to investigate the behavior of the isotope mixing state. The technique revealed that the non-mixing and the mixing states of hydrogen isotopes can be realized in plasmas. In deuterium plasmas, we also succeeded in simultaneously realizing the formation of the edge transport barrier (ETB) and the divertor detachment. It is found that resonant magnetic perturbation plays an important role in the simultaneous formation of the ETB and the detachment. Contributions to fusion reactor development from the engineering point of view, i.e. negative-ion based neutral beam injector research and the mass balance study of tritium, are also discussed. 

Recent results from deuterium experiments on the large helical device and their contribution to fusion reactor development

The tritium exhaust behavior from the Large Helical Device (LHD) was observed in the first deuterium plasma experimental campaign Tritium in the exhaust gas was monitored at the conducted by use o

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Exhaust behavior of tritium from the large helical device in the first deuterium plasma experiment

A water bubbler system that can distinguish chemical forms of tritium was proposed for long-term tritium monitoring of the exhaust gas of a large fusion test device. The characteristics and performance of the water bubbler system were evaluated under operational conditions and confirmed to be suitable for tritium monitoring. For the tritium measurements, the water bubbler system determined the tritium activity and distinguished the chemical forms of tritium. The tritium activity and chemical forms in the exhaust gas provided helpful information to understand the tritium behavior in the large fusion test device.

Determination of tritium activity and chemical forms in the exhaust gas from a large fusion test device

Tritium Balance in Large Helical Device during and after the First Deuterium Plasma Experiment Campaign

A small amount of tritium is produced by deuterium–deuterium reactions in the core plasma during deuterium plasma experiments conducted in the Large Helical Device (LHD). From the viewpoints of tritium safety handling and public acceptance, any waste gas streams arising from the vacuum vessel of the LHD and Neutral Beam Injection (NBI) system will be treated by an exhaust detritiation system (EDS). The exhaust gas stream produced during the 18th cycle of LHD plasma experiments was observed to design the EDS. The rough pumping gas streams from the LHD and NBIs produced a high flow rate of 300 Nm3/h. The vacuum exhaust gas stream from the LHD had a high hydrogen concentration more than 7% and a flow rate of less than 5 Nm3/h during plasma experiments. The regeneration gas stream of the cryopump systems had a flow rate of more than 5 Nm3/h. Based on these observations, the proposed EDS utilizes two types of tritium recovery system. One comprises two types of catalyst and dual absorbent beds. Other comprises an oxidation catalyst and polymer membrane system.

Naoyuki Suzuki

Papers

Design and commissioning of the exhaust detritiation system for the Large Helical Device

Exhaust behavior of tritium from the large helical device in the first deuterium plasma experiment

Determination of tritium activity and chemical forms in the exhaust gas from a large fusion test device

Tritium Balance in Large Helical Device during and after the First Deuterium Plasma Experiment Campaign

Observations of the Gas Stream from the Large Helical Device for the Design of an Exhaust Detritiation System