燃料粒子制御に向けた磁場閉じ込め核融合プラズマにおける 粒子排気のこれまでとこれから

本解説では，粒子制御に向けたこれまでの取り組みとして，主として粒子排気という視点に立ち，プラズマ対向壁の燃料粒子吸蔵，放出現象のほか，ダイバータによる効率的な燃料排出について筆者が行ってきた研究を中心に紹介する．また，粒子制御を行うための重要な計測手法として，中性粒子圧力計測の一つである高速イオンゲージ開発の進展についても触れる．そのほか，燃焼プラズマで想定される水素同位体，ヘリウム混合プラズマ中の同位体分離や，効率的な粒子排出を可能にする粒子制御手法についても紹介し，これからの粒子制御に向けた展望を述べる

Twenty barrel in situ pipe gun type solid hydrogen pellet injector for the Large Helical Device

A 20 barrel solid hydrogen pellet injector, which is able to inject 20 cylindrical pellets with a diameter and length of between 3.0 and 3.8 mm at the velocity of 1200 m/s, has been developed for the purpose of direct core fueling in LHD (Large Helical Device). The in situ pipe gun concept with the use of compact cryo-coolers enables stable operation as a fundamental facility in plasma experiments. The combination of the two types of pellet injection timing control modes, i.e., pre-programing mode and real-time control mode, allows the build-up and sustainment of high density plasma around the density limit. The pellet injector has demonstrated stable operation characteristics during the past three years of LHD experiments

Tritium retention characteristics of the dust in LHD after the deuterium plasma experiment

Tritium (T) retention of the dust particles collected at ten toroidal sections of the Large Helical Device (LHD) after the 21st experimental campaign, in which deuterium (DD) plasma experiments were conducted, were evaluated by an enhanced full combustion method (EFCM), and their individual T retention was characterized by a tritium imaging plate technique (TIPT) in combination with a scanning electron microscopy (SEM-EDS). A new method of fixation and storage of the dust particles containing T, called the electroconductive Resin surface Embedding Method (eREM), was successfully demonstrated for T retention characterization of the individual dust particles. From the EFCM results, total T retention in dust particles collected from all toroidal sections of LHD after the 21st campaign was found to be several hundred MBq. Much higher T retention was found in the flaky C dust particles at the toroidal section with the open helical divertor than in the other closed helical divertor sections. T retention in dust particles at the section of W-coated C divertor tiles was lower than at the other sections of C divertor tiles. Zr-Ti-V dust particles originating from in-vessel getter pump materials were found, and T retention in these particles was much lower than in the C dust particles. The results from TIPT in combination with SEM-EDS indicated not all the dust particles but a half of the dust particles retained T. Such the T retention characteristics of the dust particles could be attributed to T retention characteristics of the dust source resulting in the dust particles produced and transported, and the other is depth distribution of T in the individual dust particle thicker than T β-electron escaping range.journal articl

Experimental study of non-inductive current in Heliotron J

It is important to control non-inductive current for generation and steady-state operation of highperformance plasmas in toroidal fusion devices. Helical devices allow dynamic control of non-inductivecurrent through a wide variety of magnetic configurations. The reversal of non-inductive current consisting of bootstrap current and electron cyclotron driven current in electron cyclotron heating plasmas has been observed in a specific configuration at low density in Heliotron J device. By analyzing thenon-inductive current for normal and reversed magnetic fields, we present experimental evidence for the reversal of bootstrap current. Our experiments and calculations suggest that the reversal is caused bya positive radial electric field of about 10 kV/m. Moreover, we show that the typical electron cyclotron current drive efficiency in Heliotron J plasma is about 1.0 × 1017 AW?1m?2, which is comparable to other helical devices. We have found that the value is about 10 times lower than that of tokamak devices. This might be due to an enhanced Ohkawa effect by trapped particles

LHDヘリカル閉ダイバータのための新コンセプト真空容器内クライオソープションポンプの開発

The in-vessel cryo-sorption pump for the Closed Helical Divertor (CHD) in the Large Helical Device (LHD) has been developed at the National Institute for Fusion Science (NIFS). An organic adhesive-free bonding technique for attaching activated carbon pellets to a copper cold panel was invented, which employs the indium solder with intermediate materials. The prototype of the CHD with the newly developed cryo-sorption pump was installed in the LHD. Performance of the cryo-sorption pump was estimated in the LHD vacuum vessel. A satisfactory result of the maximum pumping speed up to 9 m3/s was obtained with one divertor module in one toroidal section (10% of the torus), which is equivalent to the required pumping speed of the CHD

Ultrahigh neutral pressures in the sub-divertor of the Large Helical Device

In the Large Helical Device (LHD) a low temperature mode (LTM) of the helical divertor was discovered. It combines particle detachment and very large sub-divertor pressures up to 1.4 Pa. During the LTM, the electron temperature in the divertor was in the range from 0.25 to 0.42 eV so that volume recombination occurred. This result is remarkable because in the stellarators LHD and Wendelstein 7-X only low sub-divertor pressures (0.03–0.3 Pa) were expected and measured up to now due the loss of pressure conservation along flux tubes by an enhanced cross-field transport. It demonstrates that the more complex, three-dimensional divertors of stellarators can achieve a similar performance with respect to particle exhaust and detachment like the geometrically simpler poloidal divertors in tokamaks even if the favorable effect of flux amplification is absent. The LTM of the helical divertor depends, however, on the magnetic configuration, i.e. on geometry. It was only observed in the inward shifted configuration with Rax = 3.55 m, but not in the more frequently studied configuration with Rax = 3.6 m. A nuclear fusion reactor based on the heliotron concept (DEMO) would benefit from the LTM by the very compact divertor configuration and the excellent performance.journal articl

Performance of ITER pressure gauges during deuterium operation in the large helical device

During deuterium campaigns on the heliotron large helical device (LHD), ITER pressure gauges with different cathode materials were used to measure the neutral pressure in the sub-divertor region. Throughout these campaigns, it was observed that the performance of a LaB6 cathode was unsatisfactory during deuterium operation. Conversely, measurements taken with pressure gauges with a ZrC cathode performed well throughout the deuterium pulses. The ITER pressure gauge with the ZrC cathode could be operated with a very high electron current of 800 µA, thus improving the lower detection limit of the neutral pressure in LHD. With this design it was also possible to avoid jumps in the ion current within strong magnetic fields, improving the accuracy of the measurement from 15% uncertainty to 5%. These features allowed very precise neutral pressure measurements to be made in a fusion device with magnetic confinement. The problems with the lifetime of the ZrC cathode reported in Mackel et al (2023 Fusion Eng. Des. 189 113439) were not relevant here, as the total runtime in the magnetic field was about 60 h, less than the expected lifetime of the cathode of 350 h.journal articl

Asymmetry in particle load on divertor tiles in different magnetic field configurations of LHD

ORCID　 0000-0002-9974-2359The asymmetry of plasma-particle load on divertor tiles at helically symmetrical positions has been investigated in the Large Helical Device (LHD). The asymmetry reverses when the toroidal magnetic field direction changes, suggesting that particle drifts cause the asymmetry. This study conducted proton orbit tracing calculations under a vacuum condition to investigate the effects of grad-B drift and curvature drift on the asymmetry. The calculations were performed for the major radius of the magnetic axis, Rax, at 3.75 m and 3.6 m configurations. The results showed that due to the effects of grad-B drift and curvature drift, a similar asymmetry to the experimentally observed one appeared in the number of protons reaching the divertor tiles. The degree of asymmetry (DOA), representing the ratio of protons reaching the symmetrical divertor tiles, increased with higher proton energy and was smaller for the Rax = 3.75 m configuration than the Rax = 3.6 m one. An analysis of the experimental data for these magnetic field configurations revealed a consistent asymmetry between the Rax = 3.75 m and the Rax = 3.6 m configuration. It was also found that the asymmetry increased with higher electron temperature and was smaller in the Rax = 3.75 m configuration than in the Rax = 3.6 m one, which was consistent with the calculations. On the other hand, the experiments showed that the DOA saturated, which was not reproduced in the calculations.journal articl

Transition between Isotope-Mixing and Nonmixing States in Hydrogen-Deuterium Mixture Plasmas

The transition between isotope-mixing and nonmixing states in hydrogen-deuterium mixture plasmas is observed in the isotope (hydrogen and deuterium) mixture plasma in the Large Helical Device. In the nonmixing state, the isotope density ratio profile is nonuniform when the beam fueling isotope species differs from the recycling isotope species and the profile varies significantly depending on the ratio of the recycling isotope species, although the electron density profile shape is unchanged. The fast transition from nonmixing state to isotope-mixing state (nearly uniform profile of isotope ion density ratio) is observed associated with the change of electron density profile from peaked to hollow profile by the pellet injection near the plasma periphery. The transition from nonmixing to isotope-mixing state strongly correlates with the increase of turbulence measurements and the transition of turbulence state from TEM to ion temperature gradient is predicted by gyrokinetic simulation

Plasma Wall Interaction of New Type of Divertor Heat Removal Component in LHD Fabricated by Advanced Multi-Step Brazing (AMSB)

ORCID　0000-0002-3744-2481A novel method, called Advanced Multi-Step Brazing, was developed to fabricate a new type of divertor heat removal component with W armor and an oxide-dispersion-strengthened copper (GlidCop®) heat sink in the initial phase of our work. Later, a new type of divertor heat removal component, which has a rectangular-shaped cooling channel with a V-shaped staggered-rib structure in the GlidCop heat sink, was developed. This new component showed an extremely high heat removal capability during a ~30 MW/m2 steady-state heat loading condition in our previous work. In this work, the new component was installed in the divertor strike position of the Large Helical Device and exposed to neutral beam injection–heated plasma discharges with 1180 shots (~8000 s) in total. Though submillimeter-scale damage, such as unipolar arc trails and microscale cracks, was identified on the W surface, the extremely high heat removal capability did not show any sign of degradation over the experimental period. On the other hand, remarkable sputtering erosion and redeposition phenomena, due to the strong influx of the divertor plasma, was confirmed on the W armor.journal articl