Design and test of a waveguide filter for reflected electromagnetic waves

A high power gyrotron system is used for plasma production and sustainment by electron cyclotron resonance heating (ECRH) in fusion reactors. An ECRH system in the Large Helical Device uses gyrotrons with megawatt output power. This megawatt electromagnetic wave transmitted through a circular corrugated waveguide is partially reflected back to the gyrotron and can be a cause of unstable gyrotron oscillations. The reflected beam may also propagate at an angle relative to the waveguide axis. To reduce the reflected beam while retaining as much of the forward-propagating gyrotron beam as possible, we aim to optimize the length of a filter consisting of a gap with absorbing walls between waveguide sections in a transmission line. A semi-analytical model is developed using diffraction theory to calculate the transmitted power of the HE11 mode as a function of filter gap distance for a wide range of beam angles, including a perfectly aligned beam to approximate the forward gyrotron beam. Before implementing this filter, we use a vector network analyzer to measure the scattering parameters for an experimental mock-up of this filter with 88.9 mm inner diameter waveguides, gap distances between 0 and 600 mm, and beam angles between 0° and 4°. These sophisticated models are used to inform the gap filter design and length optimization.journal articl

Evaluation on the operational state of turboexpanders in a helium refrigerator for nuclear fusion experimental devices using principal component analysis

ORCID　 0000-0002-1989-867XThe operational state of helium turboexpanders has been evaluated using principal component analysis (PCA). For the evaluation, two models which were single and multiple ones were utilized. The single and the multiple models assess a single turboexpander and seven turboexpanders, respectively. In the model development, we used measurement data of helium turboexpanders for a cryogenic system in the Large Helical Device. Using the two models, the changes in the turboexpanders' condition before and after the occurrence of the trouble were clarified. Consequently, the PCA is very useful to monitor the turboexpanders. Comparing the two models, the single is more practical than the multiple one.journal articl

Accident-tolerant hybrid ceramics for fusion breeding blanket

ORCID　0000-0001-8067-8732In a water-cooled ceramic breeding blanket for fusion reactors, hydrogen gas generation by steam oxidation of metallic Be compounds (i.e. neutron multiplier) in a loss-of-coolant accident (LOCA) raises major safety concerns. Li–Be hybrid ceramics has a potential to reduce hydrogen generation significantly, however, the stable compositions and structures for quaternary compositions have not been comprehensively understood. Herein, we report machine-learning based prediction, synthesis, structure, and properties of chemically stabile two-phase Li–Be–X–O hybrid ceramics. The steam exposure tests demonstrated a negligibly small H2 generation from the two-phase powder of Li2BeSiO4 and 5 at.% BeO below 1200 °C. The stability is explained by the intrinsic ionic/covalent bonding characters and little capacity for further oxidation by steam. Neutronic calculations with simplified one-dimensional model show that the two-phase hybrid cermaics has a sufficient tritium breeding capability without having metallic Be-based multiplier in the blanket. The hybrid ceramics is the first example of multi-functional oxide to breed sufficient fuel tritium with no metallic neutron multiplier, which allows a novel design of ceramic breeding blanket with enhanced safety margins during in-box LOCA.journal articl

Erratum: “Estimation of the Tritium Yields in Deuterium Fusion Plasmas Considering the Fast-Ion Velocity Distribution Function” [Plasma Fusion Res. 17, 2402023 (2022)]

ORCID　 0000-0003-3293-488Xjournal articl

Engineering design and manufacturing of the modular coil system for the quasi-axisymmetric stellarator CFQS-T

The Quasi-axisymmetric Stellarator CFQS has been constructed as an international joint project between the National Institute for Fusion Science in Japan and Southwest Jiaotong University in China. Its physical properties are as follows: toroidal periodic number m = 2, aspect ratio Ap = 4, maximum magnetic field strength Bt= 1 T, and major radius of magnetic axis in vacuum R0 = 1 m. The CFQS employs four different types of 16 modular coils to realize a quasi-axisymmetric magnetic field configuration. The CFQS plasma experiment is planned to have two stages. The first is a 0.09 T long-pulse operation with a simplified modular coil support structure referred to as “CFQS-T”. The second is a 1 T short pulse rated operation known as “CFQS” after reinforcing the modular coil system's support structure against a large electromagnetic force. This paper reports on the finalized engineering design of the modular coil system with simplified coil support structures for CFQS-T, including a validity evaluation result with finite element analysis method software ANSYS. Furthermore, the established manufacturing method of the modular coil system for CFQS-T is described. All modular coils for CFQS-T have been manufactured and its main body assembly, with modular coils and a vacuum vessel, has been completed. Currently the commissioning, including a coil energization test and magnetic field line mapping, is ongoing.journal articl

Surface modifications on unirradiated and ion-irradiated tungsten after exposed to deuterium plasma at LHD divertor-leg position

ORCID　0000-0001-5089-3642In order to clarify the divertor plasma-induced tungsten (W) surface modifications as well as the irradiation defects effect, two kinds of ITER grade W were exposed to the large helical device deuterium (D) plasma at the divertor-leg position. One was the iron (Fe) ion irradiated W to produce irradiation defects, and the other was the unirradiated W. The distributions of divertor plasma-induced surface modifications on these two kinds of W were clarified by scanning electron microscope and transmission electron microscopy. A co-deposition layer which was mainly made up of carbon (C) and Fe was formed at the private flux region (>21 mm). No significant surface change was observed at the strike point region (14–16 mm). The oxygen-enriched amorphous W structures (OEAWs) caused by plasma surface interactions were observed at the 2–14 mm (scrape-off layer region) and 20 mm. At the 2–14 mm and 20 mm, the OEAWs density on the pre-irradiated W sample is lower than that on the unirradiated W sample. On the other hand, the OEAWs size on the pre-irradiated W sample is larger than that on the unirradiated W sample at the 8–12 mm and 20 mm. While, the OEAWs size on the pre-irradiated W sample is smaller than that on the unirradiated W sample at the 2–6 mm. This study implies the possibility of forming OEAWs on the surface of W divertor. And the irradiation defects affect the density and size of OEAWs.journal articl

TALIF measurements of atomic deuterium in toroidal divertor simulator NAGDIS-T

The behavior of neutral deuterium (D) atoms is important for understanding the physics in the divertor region of nuclear fusion reactors. It is necessary to introduce an active measurement when considering the application to detached plasmas, where recombination processes dominate the processes determining the population distribution. In this work, we have developed a two-photon absorption laser-induced fluorescence (TALIF) system in the toroidal divertor simulator NAGDIS-T to measure D atomic density. The absolute D atomic density was obtained by calibrating the signal with the krypton TALIF signal. The gas pressure and power dependence of the D atomic density is shown. The D atomic density was in the range of 1.6 × 1018–1.4 × 1019 m−3, and the temperature was estimated to be <0.4 eV. The behavior of the D atoms is discussed in terms of the production process.journal articl

Investigation of island size effect on radiation distribution during attached and detached plasmas in the island divertor of W7-X

ORCID　0009-0002-8230-0121Mitigation of heat on the first wall through divertor operation is a key to a successful future fusion reactor. W7-X employs an island divertor to control the exhaust and heat load on the plasma impacting divertor plates. Increased radiation in the divertor reduces the heat load at the plasma contact point during detachment. In this paper we investigate the distribution of the radiation using an InfraRed imaging Video Bolometer (IRVB) that views the divertor region in two dimensions giving information on both the poloidal and toroidal variation of the radiation in comparison to conventional resistive bolometer arrays that typically only give poloidal variation information. Experiments were carried out using a standard magnetic configuration modified by changing control and planar coil currents to achieve three different island sizes without changing the strike line location. For each island size low and high density (ne = ∼4 and ∼ 7 x 1019/m3, respectively) plasmas were created with ∼ 2 MW of ECH input power, which correspond to attached and detached plasmas with radiated power fractions (frad) of ∼ 20–25 % and ∼ 90 %, respectively. Results indicate an increase in density led to an increase in the IRVB radiation signals as seen in the total radiated power (and frad) and a slight broadening in the signals indicating less radiation from the target locations, especially the lower right location in the IRVB field of view when compared with the corresponding thermography images. However, no noticeable difference in the IRVB radiation pattern or intensity is seen with the change of the island size.journal articl

Solution space and effective model for turbulent transport of helical plasmas

ORCID　0000-0002-2459-2392We discuss an effective transport model of magnetized turbulenthelical plasma based on the solution space of first-principle gyrokinetic simulations. If the time evolution of dynamical systems can be regarded as solution trajectories in theoretical phase space, physical phenomena in the saturated stable phase are realized in the solution space formed by these trajectories in the long time limit. Similar properties are found not only in dynamical systems but also in general physical systems with renormalization group flows (Wilson and Kogut 1974 Phys. Rep. C12 75). Therefore, if the solutions effectively form a finite dimensional solution space, the physical system can generally be represented in reduced form. Here, we try to apply this discussion to develop a transport model of turbulent plasma in first-principle gyrokinetic simulations. Based on the solution space due to the trajectory of the simulations with a certain functional form (Fujii and Nunami 2022 Plasma Fusion Res. 17 2403030), we discuss the effective structure of the objective function to represent the transport model. By evaluating the structure with fitting errors of the objective function in the model parameter space, we can determine a plausible functional form. This paper discusses a methodology for constructing such an effective transport model for helical plasmas.journal articl

In-situ calibration of charge exchange spectroscopy for spatially resolved measurements of helium-hydrogen density ratio in Wendelstein 7-X

In-situ calibration of charge exchange spectroscopy for spatially resolved measurements of helium-hydrogen density ratio is described in this paper. The helium-hydrogen density ratio in the core plasma is derived from the intensity ratio of the active helium line HeII (468.6 nm) and hydrogen line HI (656.3 nm) due to the charge exchange recombination process between fully ionized ions and neutral beam. This system is calibrated by the helium-hydrogen density ratio evaluated from HeII (656.0 nm) and HI (656.3 nm) in the recombining phase at the end of the discharge. The helium-hydrogen density ratio at the plasma center derived from the calibrated charge exchange spectroscopy is compared with the influx ratio estimated from the passive spectroscopy.journal articl