Teruya Tanaka

FLiNaBe is a promising candidate for liquid blanket materials in fusion reactors due to its chemical stability, low MHD effect, and low melting point. However, its low hydrogen solubility raises concerns regarding the permeation loss of bred tritium. While the addition of Ti powder has been proposed to increase the effective tritium solubility in FLiNaBe, the tritium release behavior from FLiNaBe mixed with Ti is not well understood. In this study, FLiNaBe samples, prepared with Ti powder under different pretreatment conditions, were irradiated by neutrons at Kyoto University Research Reactor. Subsequently the sample was heated to 600 °C or more in an Ar gas flow at Kyushu University. The released tritium was separately collected as water-soluble (T2O and TF) and non-water-soluble tritium (T2). The addition of 5 wt% as-received Ti powder significantly reduced the total tritium release, which amounted to only 60 % compared to the Ti-free FLiNaBe sample. The addition of only 0.5 wt% of pretreated Ti powder (subjected to hydrogen absorption-desorption cycles) to the 4.5 wt% as-received Ti mixture further reduced the tritium release amount to 44 % of that from Ti-free FLiNaBe. The amount of tritium released from the sample containing 5 wt% hydrogen pre-absorbed Ti was comparable to that of Ti-free FLiNaBe, indicating little tritium absorption into the Ti saturated by hydrogen. For all samples, the majority of the released tritium was water-soluble. It is suggested that tritium fluoride (TF) formed in FLiNaBe reacted with impurity metal oxides, such as BeO, and was subsequently released as T2O. A slight increase in the release of water-insoluble tritium was observed upon the addition of Ti, suggesting that Ti reduced T2O to gaseous tritium.

A 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.

We are proposing to connect a supercritical CO₂ gas turbine (SCOT) power generator to the Operational Recovery of Separated Hydrogen and Heat Inquiry-2 (Oroshhi-2) nuclear fusion primary molten salt test loop constructed at the National Institute for Fusion Science and conduct a demonstration test of the power generation principle from the world’s first nuclear fusion simulated heat output. In this paper, we describe the results of a design study of the FLiNaK/CO₂ intermediate heat exchanger and SCOT power generation system and related equipment that can be connected to the molten salt test loop. From the results of this study, it was concluded that the power generation principle verification test with 20-kW(electric)–class power generation with an efficiency of 20% is possible by the SCOT system connected to the FLiNaK test loop in Oroshhi-2 equipment via the intermediate heat exchanger.

Compatibility between tritium self-sufficiency and neutron shielding performance of self-cooled liquid blanket systems without solid neutron multiplier was investigated for the purpose of application to the conceptual helical reactor design of modified FFHR2 having the blanket space of 120 cm. The results of the neutronics calculation indicated that all of the Li/V-alloy, Flibe/V-alloy and Flibe/JLF-1 (Reduced Activation Ferritic steel) blankets are feasible for tritium breeding ability and neutron shielding performance. With the use of vanadium alloys, operation efficiency will be enhanced.

Synthetic diamond radiation detectors made of ultrahigh-purity type IIa diamond crystals were developed. The diamond crystals, which were grown by high-pressure and high-temperature synthesis, contained less nitrogen impurity than 0.1 ppm, i.e., 1/10 of conventional type IIa diamond crystals. Since the ultrahigh-purity type IIa diamond crystal was a p-type semiconductor due to a trace impurity of boron, a structure consisting of Schottky and ohmic contacts was adopted for the detectors to suppress an increase in leakage current. For the same purpose, a metal–insulator–semiconductor structure using a chemical vapor deposition diamond insulating layer was adopted for one of the detectors. Plural peaks were observed in the 5.486 MeV α energy spectra obtained by the detectors; motion of holes produced the main part of output pulses from the detectors on these measurements. One of the peaks in each spectrum was hardly influenced by light irradiation, even ultraviolet light of 370 nm (3.4 eV), compared with a peak obtained by the conventional type IIa diamond radiation detector. This experimental result indicated that there was a superior part in the crystal where deep energy levels did not exist so much. Although there was no direct evidence, influence of a distribution of the boron impurity in the crystal was expected. On the other hand, electron trapping severely occurred and resulted in loss in charge collection; an effect of the reduction in the nitrogen impurity was not confirmed on the behavior of electrons.