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Superconducting magnets in fusion reactors are subjected to a huge electromagnetic force of >100 MN/m. The magnets have to be sustained with a strong-body structure to avoid high stress and deformation. The total weight of the magnet system in the fusion reactor is estimated to be more than 20,000 tons. We applied topology optimization technique to the magnet support structure to reduce the weight of fusion reactors. Compared with the conventional design, we achieved a weight reduction of >25%. Static and seismic analyses were carried out to validate the soundness of the topology-optimized design. Consequently, the stress against the electromagnetic force in the structure was within the permissible range. It was discovered that using seismic isolation structure can adequately prevent the damage to the magnet system even when directly subjected to a massive earthquake.

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 mo...

In 2013–2014, the FliNaK/LiPb twin-loop system Operational Recovery Of Separated Hydrogen and Heat Inquiry-2 (Oroshhi-2) was constructed at National Institute for Fusion Science as a liquid blanket...

To examine the impact of nuclear transmutation in K-type and N-type thermocouples on temperature measurements in a fusion reactor, thermocouples with altered compositions were fabricated, a...

γ-ray irradiation effect on deuterium permeation through three kinds of ceramic coatings with different crystal structures has been investigated. Deuterium permeation flux through the coatings increased during γ-ray irradiation at 250–300 °C, but changes in the permeation flux were not detected at above 300 °C. The permeation flux through an erbium oxide coating increased due to crack formation during a permeation experiment at 500 °C. After deterioration of the coating, the γ-ray irradiation effect on permeation through the coating was similar to that of reduced activation ferritic/martensitic steel F82H substrate. On the other hand, an yttrium oxide coating had an amorphous structure and started crystallization at 500 °C. Although the irradiation effect before the crystallization was derived from the F82H substrate, the effect derived from the coating was detected after crystallization because the rate-determining process of permeation shifted to the coating. Compared to the results at 300 °C before and after crystallization, the irradiation effect decreased, proving that the ceramic coatings reduced the deuterium permeation as well as the γ-ray irradiation effect.

This report introduces a current status of the continuous tritium recovery test campaign of PbLi droplets in vacuum, using vacuum sieve tray (VST). The campaign aims to verify the viability of VST method at a prototype design level. The following verification are to be performed. 1) Verify the steady state extraction efficiency of tritium using a 0.6 mm nozzle. The target is greater than 80%. 2) Verify the mutual interference effects on efficiency degradation by multiple droplets by comparing 1, 4, 7, and 19 nozzles. 3) Verify the reliability of VST in continuous operation. At least 24 h of non-stop operation without efficiency degradation is the primary goal, which can be extended up to 48 h. A dedicated extraction setup is to be installed March 2019 into the Oroshhi-2 PbLi experimental loop at the National Institute for Fusion Science (NIFS) Toki-Japan, followed by a stand-alone test with a short flowing period at Kyoto University which is already finished. Deuterium (D2), instead of tritium, is used for the experiment. Extraction efficiency is calculated by comparing the D2 concentration before-drop and after-drop. The amount of released D2 is also measured by a quadrupole mass spectrometer (QMS) and verified by the concentration difference. The PbLi temperature is between 375 ℃ and 450 ℃, and the flow velocity at the nozzle is between 1.5 m s−1 and 4.5 m s−1. The final report is scheduled to be completed by the end of 2019.

Heat removal techniques of various liquid divertor concepts have been widely studied. In this study, thermal mixing characteristics of the liquid metal film-flow with locally heated on the surface under the vertical magnetic field were experimentally investigated by using various types of obstacles as a vortex generator. Especially, the efficiency of heat transport was investigated by the thermocouples installed on the bottom-wall. The velocity distributions were measured using the electric potential probe rake on the downstream of the channel. A delta-wing, a hemisphere or a cubic obstacle was used as a vortex generator installed at the center of the bottom-wall. The experiments were conducted for laminar flow region where Re = 487–1583 and in the range of N (=Ha2/Re) = 0–149 in the vertical magnetic field in the acrylic rectangular duct. GaInSn alloy was used as a working fluid. According to the comparison of heat flux distributions on the bottom-wall, the entire distributions moved towards the upstream with increasing of the strength of the vertical magnetic field for all cases. This tendency means the heat removal from the free-surface of liquid metal film-flow to the bottom-wall can be shortly achieved in case of the relatively high vertical magnetic field.

Design studies on the helical fusion reactor FFHR-c1 have progressed. The main goal of the FFHR-c1 is to demonstrate one-year steady-state sustainment of the fusion plasma with self-produced fuel and electricity. The major radius of the plasma, R, is ˜10 m and the magnetic field strength at the plasma center, B, is ˜8 T. The magnet coil arrangement is basically similar to that of the LHD. The fusion gain during the one-year operation will be ˜15 with the fusion output of ˜370 MW. Three new technologies of the high-temperature superconducting magnet coils, the cartridge-type liquid blanket, and the fusible metal pebble divertor are adopted. Including these three, we have defined 22 important issues. The strategy to efficiently address the 22 issues has been also discussed. As a result, we propose a step-by-step approach by developing the FFHR-a1 (R ˜ 2.5 m and B ˜ 4 T) and the FFHR-b1 (R ˜ 3.6 m and B ˜ 6 T), before the FFHR-c1. The FFHR-b1 is a HElical Volumetric Neutron Source (HEVNS) driven by the neutral beam injection heating, and the FFHR-a1 corresponds to the pilot HEVNS.

In our previous work, the joint between oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu–0.3 wt%Al2O3) and tungsten (W) demonstrated superior fracture strength (∼200 MPa). This joint was fabricated by the direct brazing method between W and ODS-Cu using BNi-6 (Ni–11%P) filler material without any intermediate layer. This method was named as the improved or the advanced brazing technique. In the present study, deformation and fracture behavior of the joint after the three-point bending test was investigated. At first, it was found that the crack initiation points were dominantly in the W bulk, although it was not clear that the crack initiated from grain boundary or not. Secondly, the crack propagation proceeded mostly in the W bulk but tended to deflect towards the bonding layer. These results interpret the strength of the bonding interface is superior and the present bonding technique is applicable for severe environments such as a high heat flux divertor component on the fusion reactor. Based on the above physical and technological understanding, we successfully fabricated the large scale divertor mock-up which has twenty-eight plates of W with each size of 20 × 20 × 5 mm3.

Fuel loss and environmental contamination by tritium permeation through structural materials are critical issues for the establishment of a fusion DEMO reactor. In this study, the effectivity of a chromium oxide layer formed on reduced activation ferritic/martensitic steel F82H as a tritium permeation barrier and its stability under simulated solid/liquid breeder blanket conditions have been investigated. A uniform 100-nm-thick chromium oxide layer was formed by heat treatment at 710 °C for 5 min in 50% argon-50% hydrogen mixed gas with the flow rate of 200 standard cubic centimeter per minute. After exposure to simulated solid breeder blanket conditions, an iron oxide layer and a spinel-type iron-chromium oxide layer formed. In the case of a liquid breeder blanket condition, the chromium oxide layer partly lost at 500 °C for 100 h. The chromium oxide-formed sample decreased deuterium permeation flux by a factor of up to 150. The permeation reduction efficiency deteriorated after exposure to a solid breeder blanket condition due to a change of the chromium oxide layer. However, the chromium oxide formation would play a role to reduce hydrogen isotope permeation even after reduction of the oxide layers.

The conceptual design of a helical fusion reactor was studied at the National Institute for Fusion Science in collaboration with other universities. Two types of the force free helical reactor (FFHR) are FFHR-d1 and FFHR-c1. FFHR-d1 is a self-ignition demonstration reactor that operates with a major radius of 15.6 m at a magnetic field intensity of 4.7 T. FFHR-c1 is a compact subignition reactor that aims to realize steady electrical self-sufficiency. Compared to FFHR-d1, FFHR-c1 has a magnetic field intensity of 7.3 T and a geometrical scale of 0.7. The location of the superconducting coils in both types of FFHR is based on that of the Large Helical Device (LHD). LHD has a major radius of 3.9 m. According to the design of LHD, the deformation must be within the required value to compensate for the accuracy of the magnetic field. According to this concept, the magnet support structure of LHD was fabricated using thick Type 316 stainless steel to impart sufficient rigidity. Thus, the stress of the magnet system of LHD is sufficiently below the permissible stress. In the case of FFHR, from the viewpoint of the reactor, a large access port is required for the maintenance of the in-vessel components. The mechanical design of the support structure is conceptualized by considering the basic thickness of the material and residual aperture space by referencing the mechanical analysis results. Details of the design concepts of LHD and FFHR-d1/FFHR-c1 as well as the results of mechanical analyses are introduced in this paper.

FFHR-d1A and c1 are the conceptual design of a helical fusion reactor. The positional relationship among superconducting coils, a pair of helical coils with two sets of vertical-field coils, are observed to be similar in both type of FFHR. Such a relation of coil configuration is based on the coil configuration of the Large Helical Device, which has been designed and constructed at the National Institute for Fusion Science. There is increasing demand to achieve an optimized coil configuration to anticipate improvements in plasma-confinement conditions. In this study, the structural design of FFHR based on the fundamental set of parameters of coil configuration is depicted, which satisfies the soundness of the structure. Further, the effects of the coil configuration parameters on the stress distributions are investigated. An effect of radius of curvature on a winding scheme of the helical coil is also discussed.

Various oxides and SiC have been proposed as candidate materials for electrical insulation coatings and inserts to suppress the magnetohydrodynamic (MHD) pressure drop in liquid metal fusion blanket systems. From the early stage of the MHD insulator development, the magnitudes of the radiation-induced conductivities (RICs), i.e., increases in conductivities by radiation excitation, have been evaluated on bulk and coated specimens by using several radiation sources. While the insulating performances of the oxide coated and SiC plate specimens are inferior to those of high quality and purity bulk materials due to cracks, pits, sintering additives, etc., the results indicate that the RIC phenomenon would not prevent the MHD insulators from achieving the required performances (10−3–10−2 S/m for coatings, 102 S/m for inserts) at the first wall of the blanket (500–700 °C, several kGy/s). The recent design investigation of blanket modules for DEMO reactor conditions provides more precise information for prediction of the performances in reactors and conditions to be simulated in the future irradiation damage studies. In the case of the coatings, the suppression of recoiled Li from the liquid metal breeder/coolant might be an issue to be considered in the development.

The R&D on SiC/SiC composites under the broader approach (BA) activities between Japan and the EU for fusion DEMO developed a fundamental database of mechanical (Task-1) and physical/chemical (Task-2) properties, with a primary target of the application of SiC/SiC composites as functional structure to be used in the dual coolant breeding blanket concept. This paper aims to summarize previous 10-years activities of the R&D of Japan and to provide the key deliverables toward the DEMO design. In Task-1, good creep and fatigue durability were first demonstrated. Besides, in-plane and inter-laminar strength anisotropy maps at elevated temperatures were comprehensively identified. In parallel, the irradiation effects of SiC materials were specifically determined as input parameters of the analytical model to provide for the irradiation-induced residual stresses. In Task-2, the apparent dose-dependence of the radiation-induced electrical conductivity and the indicative radiation-induced electrical degradation was identified by various irradiation sources. In addition, good gas confinement was identified. Furthermore, no accelerated corrosion for duration of 3000 h at below 1173 K was first demonstrated. With these achievements, it is suggested that the in-vessel component technology, e.g., material corrosion database development, activated corrosion product evaluation code development, compact module tests for validation of the key functions of the components, technology integration assessment for fusion nuclear tests, etc., should be further developed toward DEMO in near-term.

Deuterium permeation through a fusion-relevant ferritic steel F82H with and without erbium oxide coatings under γ-ray irradiation has been investigated in a temperature range from 300 to 700 °C. The deuterium permeation flux through F82H sample increased by γ-ray irradiation at lower temperatures below 450 °C. The irradiation effect increased with dose rate, and the percentage of the permeation flux gain might be several percent under the dose rate of a few Gy s‒1. Temperature of the F82H sample surface rose by about 0.5 °C depending on the dose rate, and so the γ-ray irradiation effect is mainly attributed to γ-heating. On the other hand, at higher temperature above 500 °C, no appreciable change of the deuterium permeation was observed. Similarly, the deuterium permeation flux through erbium oxide coated samples increased under γ-ray irradiation at lower temperatures (350‒450 °C), but no appreciable change of permeation flux through coatings was observed at higher temperatures (600‒700 °C). The coating surface temperature increased at lower sample temperatures by γ-heating. Keywords: Tritium, Permeation, γ-ray, Irradiation, F82H, Erbium oxide

We discussed with experiments the selective heating behavior of a mixture of molten salt and metal powder containing hydrogen in a single-mode cavity by 2.45 GHz microwave. The experimental device consists of two single-mode devices that intersected perpendicularly. In the common cavity at the two crossing single-mode devices, one microwave generates the maximum electric field of standing wave, and another microwave generates the maximum magnetic field of microwave. Thus, the sample located at the crossing can be selectively heated by the inputting electric field and/or the magnetic field. The frozen salt can be heated and melted by the electric field. Hydrogen was recovered from the metal powder in the molten salt by magnetic heating.

Heat transfer performance of divertor cooling device EVAPORON-4, which was developed in order to remove a high heat flux of 10 MW/m2 under low pumping power conditions, is evaluated. Uni-directional porous media used in EVAPORON-4 make it possible to promote evaporative heat transfer in microchannels of the porous medium due to its high effective thermal conductivity. In this study, we evaluate effect of soldering the uni-directional porous medium onto a grooved heat transfer surface on boiling and evaporative heat transfer performance. The heat transfer experiments prove that the boiling heat transfer performance is highly enhanced by soldering the porous medium onto the grooved surface due to the promotion of evaporation in the uni-directional porous medium and that the critical heat flux is also further increased. In particular, a high heat flux of 9.1 MW/m2, which is not critical heat flux, is achieved at 1.0 L/min and the liquid subcooling of 70 K. The pumping power characteristics also verify the evaporation promotion in the uni-directional porous medium and clarify the issues of EVAPORON-4 including the porous structure.

Sacrificial corrosion protection, a new use of Ti powder in molten salt blankets, is proposed. Redox reactions of mixed Ti powder into molten fluoride salts are predicted to suppress the TF corrosion of structural materials. Stainless Steel 430 (SS430) is immersed into FLiNaK containing HF at 823 K for 6 h. Before the immersion, the Ti powder either exposed to air or not exposed to air at all is mixed into FLiNaK. It has been found that either types of mixed Ti powder can suppress the HF corrosion of SS430.

The design of liquid metal cooling system for the liquid blanket concepts in magnetic fusion reactor requires the magnetohydrodynamics (MHD) friction loss coefficients for various pipes including pipes and bents to estimate the pressure drop of the whole blanket. The aim of this study is to reveal the MHD pressure drop of PbLi flow in a double-bended pipe. The experiments were conducted in the Oroshhi-2 loop having a super conducting magnet of 3T at NIFS in Japan. In this paper, the MHD pressure drops of the pipe under the magnetic field of 0.5T–3.0T, the ranges of Reynolds number (Re) of 2000–200,000 and Hartmann number (Ha) of 106–640 at the upstream, double-bended and downstream regions were measured. As the results, all the MHD friction loss coefficients in three regions were well correlated with the interaction parameter (Na = Ha2/Re) in the range of 0.1

Thermal and flow analyses on the cartridge-type molten salt blanket named the CARDISTRY-B (CARtridges Divided and InSerTed RadiallY – Blanket) for the compact helical fusion reactor FFHR-c1 have been carried out. As a result, it is shown that the maximum local temperature in the cartridges can be kept below 550 °C, which is an allowable temperature of the structure material, under the neutron wall loading condition of FFHR-c1. The CARDISTRY-B is composed of the tritium breeder (Breeding Blanket: BB) and the neutron shield (Shielding Blanket: SB). Both of them are toroidally segmented every two degrees. From a safety perspective, the molten salt FLiNaBe (melting point ∼305 °C) is chosen as the first option of the breeder material. In the case of FLiNaBe BB cartridge made of RAFM (Reduced Activation Ferritic/Martensitic steel), it is necessary to suppress the maximum temperature of BB cartridges to 550 °C or below. In this study, a finite volume method code considering beryllium (Be) pebbles inside a flow channel has been used to evaluate the cooling characteristics of a BB cartridge. By estimating the pressure drop through pebbles with the porous medium model, we have investigated how the temperature of BB cartridge change depends on the mass flow rate of FLiNaBe. The volumetric heating profile was modeled according to the neutronics calculation results obtained by the MCNP (Monte Carlo Neutron and Photon Transport) code.

The LHD-type helical fusion reactor FFHR has been studied to realize steady-state fusion power generation without a need for current drive and free from disruption. The conceptual design studies of FFHR are steadfastly progressing based on the presently ongoing experiments in the Large Helical Device (LHD). In order to enhance the attractive features of the base option of FFHR-d1A, which is similar to LHD, configuration optimization is being considered for FFHR-d1C. Slight modification of the helical coil trajectory gives an improved condition both for the plasma confinement and the MHD stability. In order to overcome the difficulty for construction and maintenance associated with the three-dimensional structure, innovative ideas are being explored for the superconducting magnet, divertor, and blanket. For the superconducting helical coils, the joint-winding method confirms a fast manufacturing process. The helical divertor is reexamined and practical feasibility is discussed. The maintenance method of the helical divertor and the helically-segmented breeder blanket is a serious issue and a plausible solution is proposed.

The mechanical durability and soundness of the several oxide coating materials used as the electrical insulator and tritium permeation barrier are important design parameters on an advanced liquid breeding blanket system. We tried to investigate the adhesion strength evaluation due to the peeling behavior on the MOCVD processed multilayered oxide coating (Er2O3/Y2O3) on the stainless steel (SUS) substrate before and after Cu2+ ion beam irradiation using the nano-scratch tester. The adhesion strength of the coating material was able to estimate easily from the scratch trace and scratch stress, and the nano-scratch test was suitable method to evaluate the mechanical durability and soundness. After the Cu2+ ion beam irradiation, the adhesion strength was decreased with increasing the displacement per atom (dpa). The adhesion strength degradation by the Cu2+ ion beam irradiation was caused by the embrittlement of the thicker and amorphous Fe-(Y)-O interlayer formation between Y2O3 buffer layer and SUS substrate based on the displacement damage dose effect. Keywords: Er2O3, MOCVD, Multi-layered coating, Ion beam irradiation, Adhesion strength, Nano-scratch method

Tritium handling in fusion reactors is important for economical fuel cycles and radiological safety; however, the extremely limited number of studies on tritium permeation through structural materials under irradiation conditions has been reported. In this study, γ-ray irradiation effect on tritium permeation has been investigated using three kinds of metals with different permeation properties including palladium, platinum, and austenitic stainless steel. An increase of deuterium signal on the quadrupole mass spectrometer during irradiation consisted of an increase of deuterium permeation flux and an electrical noise, and both were separately detected. The increase rate of permeation flux for all materials became lower as the test temperature increased. Dose rate dependence of the increase rate was confirmed. Oxidation of the austenitic stainless steel indicates a decrease of not only the deuterium permeation flux but also the irradiation effect.

To clarify the irradiation hardening behavior of neutron-irradiated tungsten(W), the formation behavior of rhenium(Re)-enriched cluster and W-Re precipitation with displacement damage in fission reactors environment was investigated by the atom probe tomography (APT) technique, and the results obtained by APT and microstructural observation by transmission electron microscopy (TEM) were compared. Neutron-irradiated pure W and W-10Re alloys were used for the analysis. These samples were irradiated to 0.96 dpa at 538 °C in a fast experimental reactor Joyo, and irradiated to 0.90 dpa at 500 °C in a mixed spectrum reactor HFIR. The chemical compositions of the matrix, Re clusters, and precipitates were determined by the AP analysis. The Joyo-irradiated samples contained lower levels of transmuted Re and Os, while the HFIR-irradiated samples contained higher levels of Re and Os. The Re-enriched clusters demonstrated characteristic spatial distribution related to defect cluster distribution in each irradiated specimen.

The applicability of tokamak, helical and laser fusion reactors as a volumetric fusion neutron source has been examined using the systems codes that have been utilised for the conceptual design of DEMO and commercial reactors in Japan. This study has clarified the characteristics of reactor-based volumetric neutron sources that can be designed based on the current physics and engineering basis with a reasonable running cost (∼5B Yen/year). Although the achievable neutron flux is 2–3 orders lower than that of accelerator driven neutron sources, tokamak and helical neutron sources can provide a much larger irradiation area for the test of large components. Laser neutron sources have both high operability and tritium breeding capacity. These reactor-based neutron sources also serve as an integrated test bed of the entire reactor system.

Corrosion of pure Cr and pure W in molten FLiNaK at 773 K was investigated to deeply understand the effect of the alloying elements on the corrosion mechanisms of reduced activation ferritic/martensitic (RAFM) steels. The molten salt blanket environment was simulated by hydrogen fluoride gas bubbling. Electrochemical polarization and weight change measurements were conducted to estimate corrosion rates. The oxidation reaction rate of pure Cr was extremely high (instantly reached 2.9 × 104 mm/year). The corrosion rate of W was lower than that of the RAFM steel, JLF-1 steel, and pure Fe by one order of magnitude. W was nobler than Ni, which is known as a corrosion-resistant metal to the molten fluoride salts and hydrogen fluoride. The compatibility of W coating with the first wall structure of blankets was evaluated with the local tritium breeding ratio (TBR). When a Pb neutron multiplier was used, thin W coating almost did not affect the local TBR.

The chemical behaviors of lead (Pb) based coolants in the air ingress accident of fast reactors were investigated by means of the thermodynamic considerations and the static oxidation experiments for Pb alloys at various chemical compositions. The results of the static oxidation tests for lead-bismuth (Pb-Bi) alloys indicated that Pb was depleted from the alloy due to the preferential formation of PbO in air at 773K. Bi was not involved in this oxidation procedure. Pb-Bi oxide and Bi2O3 were formed after the enrichment of Bi in the alloys due to the Pb depletion. The thickness of the oxide layer formed on liquid LBE at static condition was approximately 1mm. The oxidation rates of the alloys were much larger than that of the steels, and became larger with higher Pb concentration in the alloys. The compatibility of Pb-Bi alloys with stainless steel was worse when the Pb concentration in the alloys became low, since the dissolution type corrosion was promoted by the Bi composition in the alloy. The Pb-Li alloys were oxidized as they formed Li2PbO3 and Li2CO3. Then, Li was depleted from the alloy. These oxides of the coolant can be generated in the low temperature region of the reactor system after the increase of oxygen concentration in the coolant system in the air ingress accident.

The corrosion behaviors of the RAFM steel JLF-1 in liquid Pb-Li alloys with various Li concentrations were investigated by means of the corrosion tests in a liquid Pb, a liquid Pb-5Li alloy, a liquid Pb-17Li alloy and a liquid Pb-45Li alloy at 873 K for 750 h. The multiple oxide layer, which consisted of the porous outer layer by Fe3O4 and the compact inner layer by Cr-rich Fe-Cr-O, was formed on the steel surface during the immersion to the liquid Pb. Any oxide layer was not detected on the steel surface after the immersion to the liquid Pb-Li alloys, since the oxygen potential in the liquid alloy was lower than that for the formation of the oxide layer. Then, the dissolution type corrosion was caused on the steel surface. The Pb diffusion was detected along the boundaries of grains and subgarins in a martensitic structure of the steel. The corroded surface revealed a pebble-like microstructure by the immersion to the liquid Pb-45Li alloy. This surface microstructure was formed by the phase transformation and the preferential dissolution along the boundaries. The dissolution type corrosion could become larger when the Pb concentration in the alloy was larger, since the liquid Pb had larger solubility for Fe and Cr than the liquid Li. The corrosion in the Li rich Pb-Li alloy might be influenced by the concentration of dissolved impurities (i.e., carbon, oxygen and nitrogen) in the alloy.

A conceptual design for a helical fusion reactor is currently being undertaken by the National Institute for Fusion Science, Japan. The coil support structure is designed from the perspective of the allowable stress of the material. A continuous helical coil winding with a low temperature superconductor and a water-cooled divertor made of tungsten and copper alloy have been considered for use in this reactor, and it is defined as the basic option. Several flexible design options have also been proposed; these options solve existing issues in the basic option and they are treated as the challenging options. They can be implemented by modifying the structural components of the basic option. The structural design modifications need to be made in order for the challenging options and the mechanical soundness were investigated.

Oxidation control of candidate duct materials for liquid cooled blanket systems (i.e. SS316, SS316L, Inconel 600 and Hastelloy X) was studied to form a single composition Cr 2 O 3 layer on the surfaces for (i) making a stable and robust underlayer for ceramic coating fabrication by the metal organic decomposition (MOD) method and (ii) suppression of hydrogen permeation by the Cr 2 O 3 layer itself. In heat treatments of the bare plates at 700 °C in air of atmospheric pressure and air of reduced pressure of 5 Pa, oxidized Fe and/or Ni diffused to the outermost surfaces of the oxide layers and a single composition layer of Cr 2 O 3 could not obtained on the surfaces. However, the Cr 2 O 3 surface layers could be formed by a heat treatment of chrome plated plates at 700 °C in air of atmospheric pressure. MOD coatings were successfully fabricated on the stable Cr 2 O 3 surface layers. The preliminary evaluation of deuterium permeation through the Cr 2 O 3 covered plates showed that the permeation reduction factors (PRFs) of the Cr 2 O 3 layers would be 1–2 orders.

FLiNaBe is a promising liquid blanket material for a fusion reactor. Since hydrogen solubility for FLiNaBe is low, one concern is the permeation of bred tritium. In order to increase effective solubility of hydrogen, the addition of Ti powder was proposed. In this study, the influence of Ti addition on the behavior of tritium generated inside FLiNaBe by neutron irradiation was investigated. The samples of FLiNaBe and FLiNaBe mixed with Ti of 5 wt% and 0.5 wt% were prepared and the solid-state samples were irradiated by thermal neutrons at Kyoto University Research Reactor. The neutron-irradiated samples were heated to 700 °C in an Ar gas flow. The released tritium was collected by a water bubbler and the temporal variation of TF and HT release was observed. Initially, the most of tritium was released as HT from both FLiNaBe with and without Ti but eventually, the most of released tritium was HT after long time heating from FLiNaBe with Ti. This result suggests that the addition of Ti can decrease the release of TF very much. The amount of tritium released from FLiNaBe with Ti was lower than that from FLiNaBe without Ti. This result suggests that the addition of Ti can suppress the release rate of tritium from FLiNaBe.

For a strict control of tritium migration in fusion reactor fuel systems, tritium permeation barrier coatings have been developed for several decades. In liquid blanket concepts, corrosion of the coatings by liquid tritium breeders is a serious concern in addition to tritium permeation. Furthermore, the coatings are exposed to high-energy neutrons in an actual reactor, which would bring a synergy of irradiation and corrosion. In this study, zirconium oxide (ZrO2) coatings prepared by metal organic decomposition were exposed to static lithium-lead (Li-Pb) for 500 h at 500−600 oC after iron-ion irradiation with the damage concentration of up to 16 displacement per atom at room temperature and 500 oC. Corrosion products formed on all the exposed samples, and their amount and stability depended on the damage density and temperature during irradiation and Li-Pb exposure. Voids were formed by the irradiation and segregated in grain boundaries of the ZrO2 coating. The thicknesses of irradiated coatings decreased with damage concentration; therefore, Li-Pb corrosion of the coatings would be promoted by microstructure change caused by heavy-ion irradiation. These results suggest that the Li-Pb compatibility of the ZrO2 coatings would be degraded under irradiation environments.