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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.

Ceramic coatings have been investigated as tritium permeation barrier for several decades; however, there is little knowledge about irradiation effects on the coatings. In this study, yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steel substrates by radio frequency reactive magnetron sputtering, and their microstructures were investigated after iron-ion irradiation under various parameters with damage concentration of up to 20 displacement per atom. The formation of voids and an amorphous layer was confirmed in all the irradiated samples. In addition, it was found that irradiation-induced grain growth proceeded with damage concentration, and the increase of grain size was proportional to damage concentration. Furthermore, a layer with few voids formed between the crystallized and amorphous layers for the irradiated samples with higher incident energy or higher dose. Defect recovery and heat-driven grain growth and/or crystallization occurred for the sample irradiated at 500 °C, resulting in a larger average grain size than those irradiated at room temperature. Further consideration of irradiation effects to the coatings under fusion-relevant irradiation conditions is necessary.

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.

Point defects created by neutron/electron irradiation often determine the performance and the chemical stability and activity of oxide materials. Oxygen vacancies constitute a common point defect in metal oxides, such as erbium oxide (Er2O3), one of the candidate materials for tritium permeation barriers in fusion blanket systems. However, a systematic study of the oxygen vacancy properties in Er2O3 has been lacking. Here, the properties of isolated neutral and charged oxygen vacancies in Er2O3 are investigated by means of supercell total-energy calculations using a first-principles method based on density–functional theory. Vacancy formation energies, hydrogen–vacancy interactions, and geometry rearrangements around these point defects are investigated in detail. The characterization of the electronic structure of these point defects is established by the analysis of the spin-orbital density of states. It is found that the energetic and electronic properties of the oxygen vacancies depend on the chemical potentials of the O and Er atoms in Er2O3. The defect formation energy decreases when one hydrogen atom is trapped into the vacancy in both charged oxidizing and reduced environments. Also, the interstitial hydrogen 1s orbital forms multicenter bonds with the 5d orbital of the neighboring Er atoms, and the net charge transfer from the defected Er2O3 is 1.04e, 1.02e, and 0.80e for the defect charge states of 1−, 0 and 1+, respectively. We suggest that the H-decorated vacancy defect may be responsible for the increase of the self-healing properties of Er2O3.

Tritium permeation through and corrosion of structural materials are critical issues in fusion reactor liquid lithium-lead blanket concepts from the viewpoints of an efficient fuel cycle, higher operation rates, and radiological safety. In this study, lithium-lead compatibility of ceramic coatings has been investigated for the development of tritium permeation barriers with corrosion protection. Erbium oxide, yttrium oxide and zirconium oxide coatings were fabricated by a metal organic decomposition method on reduced activation ferritic/martensitic steel F82H substrates. Corrosion and delamination of the coatings were accelerated under a higher oxygen concentration. Under a lower oxygen concentration, zirconium oxide coatings had the best lithium-lead compatibility among three coating materials from surface and cross-sectional observations. However, the zirconium oxide coating after lithium-lead immersion at 550 °C for 500 h showed higher deuterium permeability in comparison to the sample without immersion. Formation of a chromium oxide layer on the surface of the substrate before fabricating the coatings drastically improved the lithium-lead compatibility of erbium oxide and yttrium oxide coatings. Degradation of the coatings was mainly caused by corrosion and delamination depending on immersion temperature, test duration, and impurity concentration.

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

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

Tritium permeation barrier coatings have been investigated for several decades to control tritium migration in fusion reactor fuel systems. In liquid lithium-lead blanket concepts, mitigation of not only tritium permeation through but also corrosion of structural materials is seriously required. In this study, for the development of a multifunctional coating, Er2O3-ZrO2 multi-layer coatings were prepared on reduced activation ferritic/martensitic steel substrates by metal organic decomposition. The deuterium permeability of the coated sample was three orders of magnitude lower than that of uncoated substrate in the first permeation experiment at 400 °C, and the high permeation reduction was kept in the temperature range of 400–650 °C. The coatings remained all over the sample surface after Li-Pb immersion experiments at 500–600 °C for 500 h, but were damaged at 600 °C. No reduction in coating thickness was confirmed after immersion at 550 °C by cross-sectional observation, suggesting that the multi-layer coatings may be applicable to the liquid lithium-lead blankets.

Tritium permeation through structural materials is one of critical issues in liquid lithium-lead blanket concepts from the viewpoints of an efficient fuel cycle and radiological safety. Metal oxide coatings have been investigated as tritium permeation barrier and showed high permeation reduction factors. For the application to DEMO reactors, however, corrosion of the coatings by blanket materials is an unavoidable concern. This paper focuses on preparation of three metal oxide, erbium oxide, yttrium oxide, and zirconium oxide coatings by a liquid phase method and comparison of their properties in terms of hydrogen isotope permeability as well as lithium-lead compatibility. The deuterium permeation behavior of the erbium oxide and yttrium oxide coatings was similar, while the zirconium oxide showed a decrease of the permeation flux by further crystallization at lower temperature than the others. The zirconium oxide coating showed the best lithium-lead compatibility among three oxides at up to 600 °C. Deterioration of the coatings after static lithium-lead exposure would be caused by delamination and corrosion. Delamination of the coating would be prevented to control the coating-substrate interface. Corrosion of the coatings by formation of ternary oxides or reduction will be the main issue in lithium-lead compatibility at high temperatures.

Erbium oxide has been investigated as a tritium permeation barrier material in a D-T fusion reactor fuel system for more than a decade. Erbium oxide normally forms the cubic phase; however, the uncommon monoclinic phase was formed under ion irradiation or high pressure conditions. In this study, the monoclinic erbium oxide coatings were prepared to examine their microstructure and deuterium permeability with ion-irradiation effect. The monoclinic phase coatings with a preferred orientation showed 1–2 orders of magnitude higher permeability at 300–400 °C than the cubic phase coating reported previously due to the smaller grain with uncrystallized region. After the permeation measurement at 600 °C, the permeability drastically decreased due to the phase transformation to the cubic phase and a change in grain structure from columnar to granular. An Fe-ion irradiated coating with the damage concentration of 0.05 dpa showed the decrease in the permeability at 500 °C, indicating the irradiation damage to the grain structure may accelerate the grain growth at lower temperature, while the phase transformation would occur at 600 °C because the diffusivity did not change much at 500 °C and drastically decreased at 600 °C.

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 evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W. Keywords: He in damaged W, Hydrogen isotope retention behavior, Damaged W, Fusion

For the establishment of liquid tritium breeding concepts, static lithium-lead corrosion tests for single-layer erbium oxide coatings and erbium oxide-metal multi-layer coatings were carried out, followed by deuterium permeation measurements. Grain boundary corrosion of erbium oxide coatings was confirmed by the static immersion tests at 550 and 600 °C. An erbium oxide-iron two-layer coating sustained its structure after lithium-lead immersion at 600 °C for up to 3000 h. The results of gas-driven deuterium permeation measurements for multi-layer coatings immersed at 550 and 600 °C indicated that crystallization and grain growth of erbium oxide would sufficiently occur during the immersion at 600 °C, and then the sample showed lower permeabilities in the first measurements at lower temperature. On the other hand, permeation reduction factors of the sample immersed at 550 °C were estimated to be 100‒200 in the temperature range of 400‒550 °C. A corrosion layer formed on the coating surface might work as an additional diffusion barrier after the permeation test at 600 °C. Keywords: Lithium-lead, Tritium permeation barrier, Corrosion, Erbium oxide

The estimation of tritium permeation rate through the plasma facing wall into coolant is required to discuss tritium balance in a D-T fusion plant, to design tritium recovery system and to perform safety assessments. In this work, tritium permeation rates in the blanket first wall and the divertor were estimated by numerical analysis for simplified multi-layer structures with considering the temperature distribution in recent JA DEMO condition. The permeation rate in the blanket first wall, which was a double layer consisting of tungsten and F82H, was estimated to be 0.69 g/day. The permeation rate in the divertor, which was a triple layer consisting of tungsten, copper and copper alloy or F82H, was estimated to be 0.013 g/day. When the permeation rate in tritium breeding region in the blanket can be reduced by three orders of magnitude due to a permeation barrier, total tritium permeation rate in the blanket and the divertor was estimated to be 0.71 g/day.

To evaluate the damage depth profile effect on hydrogen isotope retention in tungsten (W), combination usage of 0.8 MeV and 6.0 MeV Fe ions were implanted into W with the damage concentrations between 0.03 and 0.1 dpa. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed with the flux of 1.0 × 1018 D+ m−2 s−1 up to the fluence of 1.0 × 1022 D+ m−2, and the D retention behavior was evaluated by thermal desorption spectroscopy (TDS). The experimental results indicated that 6.0 MeV Fe ion irradiation would introduce vacancies and voids into bulk that were clearly controlled by the damage concentration, and the voids would become the most stable D trapping sites. It was found that D de-trapping from irradiation defects at lower temperature would be enhanced by the accumulation of defect near the surface due to 0.8 MeV Fe ion irradiation.

Ultrafine lithium titanate (Li 2 TiO 3 ) powder was synthesized by hydrothermal method. The phase formation and transition condition among α, β, and γ-Li 2 TiO 3 were discussed. XRD and ICP-AES showed the single α-phase was formed at 180 °C with 2 h hydrothermal reaction, and it transited into β-phase at 400 °C. SEM observation and EDS analysis confirmed the dissolution of TiO 2 and the formation of α-Li 2 TiO 3 proceeded simultaneously with preferable growth direction of (-133) lattice. During the phase transition, the powder maintained the small crystallite, which facilitated the fabrication of Li 2 TiO 3 bulk with small grain size. After the Ar + irradiation, the surface region to the depth of 3 μm of Li 2 TiO 3 ceramic was affected, where the decrease of crystallization and disturbance of short-range order were confirmed by GIXRD and Raman spectroscopy. In spite of the structure change at the surface area, the ceramic bulk maintained the same.

Influence of mixed material layer formation on hydrogen isotope retention in W exposed to 2014 Large Helical Device (LHD) experiment campaign was evaluated by Thermal Desorption Spectroscopy (TDS), X-ray Photoelectron Spectroscopy (XPS), Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM). It was found that a lot of hydrogen isotopes were trapped by the carbon-dominated mixed-material layer deposited on the plasma facing materials. Most of He was also trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K, respectively. However, the hydrogen retention behavior for erosion dominated area was clearly different from those for deposition dominated area and typical Plasma Wall Interaction (PWI) area, where He bubbles were introduced near the sample surface, leading to the introduction of various types of trapping sites in W.

Hydrogen isotope migration behaviors in ceramics-metal multilayer coatings have been elucidated for a further improvement of fueling system and radiological safety. Erbium oxide (Er2O3) coatings were fabricated by filtered vacuum arc deposition (VAD) on reduced activation ferritic/martensitic steel substrates. An iron (Fe) layer was fabricated by radio-frequency magnetron sputtering or covered with an Fe foil on the Er2O3 coating. An Er2O3-Fe-Er2O3 three-layer coating was also fabricated by the VAD on the Er2O3-Fe coating. The grain boundary diffusion model was constructed based on the results of grain structure analysis and deuterium depth profile for the Er2O3 single-layer coating from our previous works. The Er2O3-Fe two-layer coatings with different Fe layer thickness showed no significant difference in deuterium permeability. The Er2O3-Fe-Er2O3 three-layer coating showed a PRF of up to 104 due to a contribution of two diffusion barriers of the inner and outer Er2O3 layers. In addition, the three-layer coating had three times higher deuterium concentration than the Er2O3-Fe coating, although the concentration will be negligibly small from the perspective of tritium inventory in the fuel system.

Y 2 O 3 coatings containing Cr 3 C 2 nanoparticles have been prepared on reduced activation ferritic/martensitic steel F82H substrates by metal organic decomposition with dip-coating technique, and their deuterium permeation behaviors have been investigated. The surface condition of the samples varied with pre-treatment process: fabrication of Y 2 O 3 coatings without the Cr 3 C 2 nanoparticles or surface oxidation of F82H. In the coating without pre-treatment, peeling of agglomerated nanoparticles and oxidation of the substrate at the peeled regions were observed. Pores were also observed in the coatings with pre-treatment; however, iron oxide was not formed. The deuterium permeation flux of the coating fabricated without pre-treatment was lower than that of uncoated F82H substrate by a factor of 100 even after crystallization at 700 °C. In the case of the sample with pre-treatment, the permeation flux was by a factor of 100 lower than that of uncoated substrate before crystallization, and the factor increased to approximately 1000 after crystallization. Dispersion of the nanoparticles and securement of surface coverage of the coating are key factors to establish high-performance tritium permeation barrier.

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.

Deuterium (D) retention behavior for the sequential 6 MeV iron (Fe) and 10 keV carbon (C) implanted tungsten (W) were evaluated by thermal desorption spectroscopy (TDS) and β-ray-induced X-ray spectroscopy (BIXS) to understand the synergetic effect of defect formation and C existence on D retention behavior for W under various damage distribution profiles. The experimental results indicated that retention of D trapped by dislocation loops was controlled by 10 keV C + implantation. The D retention was reduced in the sequential Fe 2+ − C + implanted W with higher C + fluence in comparison to that with lower C + fluence due to the formation of C-W layer which suppressed D diffusion toward the bulk and dense defects at the surface which reduce effective D diffusion coefficient. On the other hand, the amount of D trapped by the defects in the deeper region than C + implantation region (50 nm) was increased due to the formation of dense defects by 6 MeV Fe 2+ implantation within the depth of 1.5 μm.

Li 2 TiO 3 is a candidate material for tritium breeding in the future nuclear fusion reactor. In this study, Li 2 TiO 3 powder was synthesized by ultrasonic-assisted solution combustion synthesis (USCS) in a single step. The ultrasonic transducer with the power of 1000 W was introduced in the synthesis process. The crystallite size of Li 2 TiO 3 powder prepared by utilization of ultrasonic power is significantly decreased to ∼5.0 nm, while the one obtained without ultrasonic power is 20.0 nm. Li 2 TiO 3 ceramic sintered from USCS powder at 800 °C exhibits the small grain size of 330 nm and the open pores size of 140 nm. The crush load of the ceramic reaches 37.2 N although the structure is porous. Compared with the ceramic prepared by solid-state reaction and conventional solution combustion synthesis, USCS sample has a higher conductivity of 2.0 × 10 −6 S m −1 at room temperature, indicating the lower tritium diffusion barrier in the ceramic.

The annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D2 spectra for Fe damaged W with lower defect concentr...

In this study, we measured deuterium (D) gas-driven permeation through tungsten (W) foils that had been pre-damaged by helium ions (He + ). The goal of this work was to determine how ion-induced damage affects hydrogen isotope permeation. At 873 K, the D permeability for W irradiated by 3.0 keV He + was approximately one order of magnitude lower than that for un-damaged W. This difference diminished with increasing temperature. Even after heating to 1173 K, the permeability returned to less than half of the value measured for un-damaged W. We propose that this is due to nucleation of He bubbles near the surface which potentially serve as a barrier to diffusion deeper into the bulk. Exposure at higher temperatures shows that the D permeability and diffusion coefficients return to levels observed for undamaged material. It is possible that these effects are linked to annealing of defects introduced by ion damage, and whether the defects are stabilized by the presence of trapped He.

All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trap...

The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle & Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium two-component system. In the representative recent DEMO condition, the following tritium permeation\ rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.

For the realization of a sufficient fuel cycle in a fusion reactor, tritium permeation barriers have been studied mainly using ceramics coatings to suppress tritium leakage through structural materials. In this study, the first trial of fabrication and characterization of yttria-stabilized zirconia (YSZ) coatings has been carried out. The YSZ coatings containing 4 mol% Y2O3 were deposited on reduced activation ferritic/martensitic steels by radio-frequency magnetron sputtering devices in two different deposition modes. The coating fabricated in the metallic mode showed a lower density with many cracks, and deteriorated after annealing and deuterium permeation tests at 600 °C. On the other hand, the coating fabricated in the poison mode showed relatively higher density, and did not deteriorate from the result of the deuterium permeation measurements at up to 600 °C. These results suggest that the YSZ coating deposited in the poison mode would be applied as a tritium permeation barrier.

For the realization of liquid metal blanket systems, fabrication of a ceramic-metal multi-layer coating has been proposed to obtain both electrical insulation for the mitigation of magneto-hydrodynamic pressure drop and compatibility with liquid metals. The corrosion protection layer using a metal foil needs to be joined to the ceramic coating to ensure integrity in the flowing environment. Besides, the ceramic coating should be as thin as possible in terms of thermal conduction. In this study, we fabricated ceramic-iron joint coatings using iron foils and thin ceramic coatings by hot pressing. Three kinds of the ceramic coating with different crystallization states were used for joining to compare adhesion to the ceramic coating. The coating before and after joining fulfilled the requirements for electric insulation. The joint coating showed a sufficient adhesion strength to withstand the stress by liquid metal flow due to formation of an iron oxide layer between the iron foil and the ceramic coating. Although the thermal conductivity of the joint coating slightly decreased due to the heat resistance of the interface between the iron foil and the ceramic coating, the joint coating is a promising functional coating for liquid metal blanket systems.

Tritium permeation through structural materials is one of the critical issues in a fusion reactor from the viewpoints of an efficient fuel cycle and radiological safety. Ceramic coatings have been studied as tritium permeation barriers and shown sufficient reduction of hydrogen isotope permeation. In high thermal efficiency liquid blanket concepts, however, not only tritium permeation but also corrosion of the structural materials by liquid tritium breeders such as lithium-lead (Li-Pb) is a serious concern. This study focuses on the fabrication of two kinds of yttria-stabilized zirconia (YSZ) coatings by metal organic decomposition and characterization through several tests such as deuterium permeation and static Li-Pb exposure. The 4 mol%-Y2O3-containing YSZ coating of less than 100 nm in thickness showed three orders of magnitude lower deuterium permeation flux than uncoated substrate after crystallization and grain growth, and the high permeation reduction performance maintained through the thermal cycle tests. The 10 mol%-Y2O3-containing YSZ coating sample showed a high permeation flux due to the defects formed during the permeation measurement. Both coatings showed compatibility with Li-Pb. A phase transformation was confirmed in the 4 mol%-Y2O3-containing YSZ coating after bending tests, suggesting that the coating sample achieved a self-repairing property. These results indicate that the 4 mol%-Y2O3-containing YSZ coating would apply to the tritium permeation barrier.

Hydrogen addition to a sweep gas of a solid breeder blanket module has been proposed to enhance tritium recovery from the surface of the breeders. However, the influence of hydrogen addition on the bred tritium behavior is not understood completely. Tritium behavior in the simplified blanket module of Li 2 TiO 3 pebbles was numerically calculated considering diffusion in the grain bulk, surface reactions on the grain surface and permeation through the cooling pipe. Although a partial pressure of T 2 increases with increasing a partial pressure of H 2 in the sweep gas, it was estimated that tritium permeation rate to the cooling water decreases. Additionally, the release duration of water vapor generated by the reaction of the pebbles and hydrogen is shortened with increasing a partial pressure of H 2 . Tritium inventory in the grain bulk and the grain surface occupies 99.6 % of total tritium inventory in the blanket module.

Yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by metal organic decomposition with a dip-coating technique, and their deuterium permeation behaviors were investigated. The microstructure of the coatings varied with heat-treatment temperature: amorphous at 670 oC (amorphous coating) and crystallized at 700 oC (crystallized coating). Deuterium permeation flux of the amorphous coating was lower than the uncoated steel by a factor of 5 at 500 oC, while that of the crystallized coating was lower by a factor of around 100 at 400‒550 oC. The permeation fluxes of both coatings were drastically decreased during the measurements at higher temperatures by a factor of up to 790 for the amorphous coating and 1000 for the crystallized one, indicating a microstructure modification occurred by an effect of test temperature with hydrogen flux. Temperature dependence of deuterium diffusivity in the coatings suggests that the decrease of the permeation flux has been derived from a decrease of the diffusivity. Characteristic permeation behaviors were observed with different annealing conditions; however, they can be interpreted using the permeation mechanism clarified in the previous erbium oxide coating studies.

Usage of CuO and water bubbler is one of the conventional and convenient methods for tritium recovery. In present work, influence of CuO particle size and diameter of reaction tubing on the tritium recovery was evaluated. Reaction rate constant of tritium with CuO particle has been calculated by the combination of experimental results and a simulation code. Then, these results were applied for exploring the dependence of reaction tubing length on tritium conversion ratio. The results showed that the surface area of CuO has a great influence on the oxidation rate constant. The frequency factor of the reaction would be approximately doubled by reducing the CuO particle size from 1.0 mm to 0.2 mm. Cross section of reaction tubing mainly affected on the duration of tritium at the temperature below 600 K. Reaction tubing with length of 1 m at temperature of 600 K would be suitable for keeping the tritium conversion ratio above 99.9%. The length of reaction tubing can be reduced by using the smaller CuO particle or increasing the CuO temperature.

Effects of annealing after/under iron (Fe) ion irradiation on deuterium (D) retention behavior in tungsten (W) were studied. The D2 TDS spectra as a function of heating temperature for 0.1 dpa damaged W showed that the D retention was clearly decreased as the annealing temperature was increased. In particular, the desorption of D trapped by voids was largely reduced by annealing at 1173 K. The TEM observation indicated that the size of dislocation loops was clearly grown, and its density was decreased by the annealing above 573 K. After annealing at 1173 K, almost all the dislocation loops were recovered. The results of positron annihilation spectroscopy suggested that the density of vacancy-type defects such as voids, was decreased as the annealing temperature was increased, while its size was increased, indicating that the D retention was reduced by the recovery of the voids. Furthermore, it was found that the desorption temperature of D trapped by the voids for damaged W above 0.3 dpa was shifted toward higher temperature side. These results lead to a conclusion that the D retention behavior is controlled by defect density. The D retention in the samples annealed during irradiation was less than that annealed after irradiation. This result shows that defects would be quickly annihilated before stabilization by annealing during irradiation.

Pure tungsten (W) samples were respectively installed on the two positions of typical plasma wall interaction area (PI) and erosion dominated position (ER) on the first wall in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan. After the experiment campaign in 2014, these samples were picked up and thermal desorption spectroscopy (TDS) were applied to evaluate their desorption behavior of He which was implanted by He discharge. It was found that a carbon-dominated mixed-material layer with impurities, such as Fe, Cr, Mo, O and N, etc., was deposited layer by layer during the plasma exposure on the PI sample. On the other hand, W-C mixed layer was formed on the ER sample. The results showed that a large amount of He was trapped in the samples on both PI and ER, and the total He retention for ER is about twice as large as that of PI. He was trapped in various types of trapping sites in the ER sample and their desorption peaks were located at temperatures of about 425 K, 755 K, 1130 K and 1630 K. For PI sample, most of He was trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K. The additional 3.0 keV helium ion (He+) implantation was performed for several samples to investigate the He retention characteristics in these samples and it was found that no additional desorption stage was found. These results suggested that the He discharge history and its deposition on the W plasma facing wall would affect He desorption behavior of W.

Tritium permeation through structural materials in fusion reactor blanket systems is a critical issue from the viewpoints of fuel loss and radiological safety. Ceramic coatings have been investigated to prevent tritium permeation; however, corrosion of the coatings by blanket materials, especially corrosive liquid tritium breeders such as lithium-lead is serious. In our previous study, the improvement of permeation reduction performance using an erbium oxide (Er2O3)-zirconium oxide (ZrO2) two-layer coating was confirmed, but it did not show substantial corrosion resistance. In this study, various multi-layer coatings were fabricated by metal organic decomposition and then exposed to lithium-lead under static conditions to investigate precise corrosion behaviors of multi-layer coatings. After lithium-lead exposure for 1000 h at 600 °C, the outermost Er2O3 layer almost disappeared, while the outermost ZrO2 layer remained, indicating that ZrO2 is more suitable as the outermost layer. However, many cracks and peelings were observed on the outermost ZrO2 layer in the cases of the samples having four coating interfaces. The optimization of layer combination and the control of adhesion between the coatings are required to reduce the degradation of multi-layer coatings.

Tritium permeation and retention are serious problems in D-T fusion reactors from the viewpoint of fuel efficiency and radiological safety. Functional ceramic coatings have been intensively studied for the development of tritium permeation barriers for several decades, while reports about tritium retention in the ceramic coatings are scarce. Moreover, irradiation may affect tritium retention in fusion materials, which is important to precisely evaluate tritium inventory in the reactor. In this study, the gamma-ray irradiation effect on deuterium retention in reduced activation ferritic/martensitic steel and three kinds of ceramic coatings were investigated through deuterium exposure, gamma-ray irradiation using cobalt-60 gamma-ray sources and deuterium depth profile measurements. The amount of deuterium retention in yttrium oxide, silicon carbide, and zirconium oxide coatings decreased after the irradiation in the dose rate of 2.43 Gy s−1, while no clear change in the retention was observed at the lower dose rate. From these results, the gamma-irradiation effect on deuterium retention would have a threshold dose rate. Diffusion and desorption of deuterium would be accelerated by excitation of deuterium via energy transfer from electrons generated by Compton scattering.

In a fusion reactor, strict control of tritium is required from the viewpoints of sufficient fuel balance and radiological hazard. Fabrication of tritium-impermeable ceramic coatings is a promising solution, and ceramic coatings have been investigated for several decades. In recent years, several irradiation tests have been conducted for the ceramic coatings to understand irradiation effects on their characteristics and functions. However, there are few experimental reports on multi-irradiation tests of neutrons and alpha particles produced by nuclear reactions and radioactive decay of tritium. In this study, yttrium oxide coatings were fabricated by reactive and non-reactive radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated after heavy ion irradiation and helium implantation tests simulating irradiation damage by neutrons and generated alpha particles. All irradiated samples showed the formation of an amorphous layer and a nanocrystalline layer or region. In particular, an iron-helium-simultaneous-irradiated sample contained voids including nanometer-size spheres, indicating the formation of helium bubbles. The irradiated samples showed grain growth and/or crystallization at a higher temperature than the unirradiated samples during deuterium permeation tests. That suggests that the voids and the helium bubbles prevented structural change at low temperatures. Moreover, the helium bubbles tended to be more stable than the voids. On the other hand, nickel-single and nickel-helium-sequential-irradiated coatings deposited by non-reactive sputtering showed few voids and helium bubbles, resulting in a different behavior of deuterium permeation.

We report the fabrication and characterization of nanocrystalline Er 2 O 3 thin films, as well as the calculations of interstitial deuterium (D) behaviors in bulk Er 2 O 3 using first-principles calculations based on density-functional theory. Our results show that the prepared Er 2 O 3 thin films possess a regular cubic phase with a phase structure consisting mainly of bulk and grain boundaries. Transport of hydrogen isotopes in bulk is predicted to occur by translocation between empty interstitials ( e.g. , tetrahedral or octahedral sites), and involves hopping from one empty interstitial to another. The rate-determining process of diffusion and permeation is hydrogen isotope movement from tetrahedral site to octahedral site. Transport of hydrogen isotopes within grain boundaries is predicted to be dominant based on the comparison of D permeability or permeation activation energy between experiments and DFT calculations. At a typical temperature of 873 K, the solubility, diffusion coefficient, and permeability of D in bulk Er 2 O 3 are predicted to be 2.94 × 10 −15 mol m −3 , 3.56 × 10 −10 m 2 s −1 , and 3.68 × 10 −27 mol m −1 s −1 Pa −1/2 , respectively.

The effect of heating temperature on deuterium (D) retention behavior for helium (He+) / carbon (C+) implanted tungsten (W) was studied. It was found that D retention behavior for He+ implanted W was not limited by the size of the He bubbles. The microstructure observation showed that the large helium bubbles were formed near the surface for He+ implanted W at 1173 K, suggesting that the D retention was reduced by the growth of the helium bubbles. In addition, to evaluate the effect of implantation ion species at high temperature, D retention behavior for He+ implanted W at 1173 K was compared with that for C+ implanted W at 673 K. It is concluded that the D retention depends on ion species, which makes different kinds of damages like He bubbles for He+ implantation and vacancy-ion complex (voids) for C+ implantation.

Dynamics of tritium recovery using CuO catalyst and water bubbler was studied as a function of gas flow rate and CuO temperature. The rate constant of tritiated water formation by CuO catalyst at t...

Tritium permeation through structural materials is a critical issue in fusion reactors from the viewpoints of sufficient fuel balance and radiological hazard. Ceramic coatings have been investigated as tritium permeation barrier for several decades; however, irradiation effects of the coatings on permeation are not elucidated. In this work, yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated before and after iron-ion irradiation at different temperatures. An as-deposited coating had a columnar structure and transformed into a granular one after annealing. An amorphous layer formed near the coating-substrate interface of irradiated coatings, and its thickness became thinner with increasing irradiation temperature. Voids of approximately 20 nm in diameter also formed in the irradiated coatings. Deuterium permeation flux of the sample irradiated to 1 dpa at room temperature was the lowest among the unirradiated and irradiated samples, and a permeation reduction factor indicated up to 390. The amorphous layer disappeared after deuterium permeation measurements due to damage recovery, while the voids remained and aggregated. The irradiation damage would accelerate nucleation of the crystal, resulting in a decrease of the permeation flux.

Erbium oxide (Er2O3) is a candidate material for tritium permeation barriers (TPB) applied to fusion blanket systems. To permeate through a TPB coating on a metal substrate, a hydrogen isotope must first adsorb on the surface, and then penetrate from the surface into the interior. In this work, Er2O3 coatings were fabricated by arc source assisted deposition and annealing at 973 K in vacuum ( e ven at high temperature.

Tritium permeation barrier has been developed for mitigating fuel loss and radiological concern at a fuel breeding/recovery system in a D-T fusion reactor. Recent research effort has been dedicated to erbium oxide coatings, and various hydrogen permeation behaviors except for irradiation effects have been elucidated. In this study, irradiation effects on deuterium permeation through erbium oxide coatings have been investigated by iron-ion irradiation at elevated temperature followed by deuterium gas-driven permeation experiments. The coatings deposited on reduced activation ferritic steel substrates with displacement damages of 0.01–1 dpa showed one or two orders of magnitude different permeabilities at 300–500 °C; however, the permeabilities became comparable and lower than that of unirradiated at 550–700 °C, indicating the grain growth and the formation of grain boundaries with a lower permeability. Cross-sectional transmission electron microscopy with selected-area electron diffraction for the coatings before and after the permeation experiments indicated the formation of a defect-accumulated region. The stability of the region strongly depends on the irradiation condition: damage concentration and annealing time, resulting in the difference of the permeability and diffusivity in the lower temperature range.

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

Tritium permeation through erbium oxide coatings has been modeled on the basis of experimental results. Permeation models were constructed step-by-step by the introduction of the following predomin...

Er 2 O 3 coating for tritium permeation barrier has been fabricated on steel substrates by a filtered arc deposition method at room temperature and 973 K. Thermal expansion of the oxide layer and the substrate induced peel-off of the coating. The non-crystalline layer is thought to play a role in forming a uniform surface coating. Five cycles of permeation measurements at 773-973 K resulted in no degradation of the coating. Different permeation behaviors are seen between degassing for 12 h at 873 K and at room temperature. Low hydrogen background following degassing at 873 K helps detect the transition to deuterium permeation. The permeation flux following different degassing conditions eventually approached comparable levels.

Deuterium (D) retention behaviours for 14 MeV neutron irradiated tungsten (W) and fission neutron irradiated W were evaluated by thermal desorption spectroscopy (TDS) to elucidate the correlation between D retention and defect formation by different energy distributions of neutrons in W at the initial stage of fusion reactor operation. These results were compared with that for Fe2+ irradiated W with various damage concentrations. Although dense vacancies and voids within the shallow region near the surface were introduced by Fe2+ irradiation, single vacancies with low concentration were distributed throughout the sample for 14 MeV neutron irradiated W. Only the dislocation loops were introduced by fission neutron irradiation at low neutron fluence. The desorption peak of D for fission neutron irradiated W was concentrated at low temperature region less than 550 K, but that for 14 MeV neutron irradiated W was extended toward the higher temperature side due to D trapping by vacancies. It can be said that the neutron energy distribution could have a large impact on irradiation defect formation and the D retention behaviour.

Temperature dependence on deuterium (D) retention for He+ implanted tungsten (W) was studied by thermal desorption spectroscopy (TDS) to evaluate the tritium retention behavior in W. The activation energies were evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code and found to be 0.55 eV, 0.65 eV, 0.80 eV and 1.00 eV. The heating scenarios clearly control the D retention behavior and, dense and large He bubbles could work as a D diffusion barrier toward the bulk, leading to D retention enhancement at lower temperature of less than 430 K, even if the damage was introduced by He+ implantation. By comparing the D retention for W, W with carbon deposit and tungsten carbide (WC), the dense carbon layer on the surface enhances the dynamic re-emission of D as hydrocarbons, and induces the reduction of D retention. However, by He+ implantation, the D retention was increased for all the samples.