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Neutron-irradiated tungsten (W) samples were exposed to helium (He)–seeded deuterium (D) plasmas using a linear plasma device called Tritium Plasma Experiment in order to investigate the synergetic...

Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET–ITER-like wall (ILW) during two campaigns in 2011–2012 (first campaign, ILW-1) and 2015–2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470°C to 520°C, which was higher than measured after ILW-1: 370°C to 430°C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns.

The deuterium (D) permeation behavior for 1 displacement per atom Fe2+ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal des...

For deuterium–tritium fusion reactor, the fuel of tritium is not available naturally. Tritium should be produced by the reaction of Li (n, ɑ) T. Li4SiO4 is one of promising candidates due to its high lithium density. In present work, effects of tritium release behavior from Li4SiO4 have been investigated and compared. The main release peak at lower temperature was attributed to tritium located on/near the grain surface and chemical adsorption on pore surface. The small peak at higher temperature was designated as tritium diffusion from grain including lattice and boundary. The tritium release behavior was simulated by diffusion model and surface reaction model. Tritium behavior is controlled by both diffusion and surface reaction for different stages. Based on the desorption theory, the kinetic parameters were obtained. The effects on tritium release behavior were studied. The porosity has significant effects on tritium release. Higher porosity has large specific surface providing more activation sites and fast channels for tritium release which results in lower temperature of tritium release. The main tritium release form was considered as tritium water. The release peak of deuterium moved towards lower temperature after adsorbing water vapor for 1800 h. It was demonstrated that adsorbing water was beneficial for tritium release due to isotope exchange. Compared the effects of porosity and water vapor, higher porosity has larger effects on tritium release in the present work. Based on the work, the tritium release is controlled by both diffusion and surface reaction at different stage and affected by comprehensive effects including porosity and adsorption water.

In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.

We investigated deuterium (D) retention in three W samples irradiated with MeV neutrons at high damage level of 0.39 ~ 0.74 displacements per atom (dpa) at high temperatures, 894 K, 1074 K and 1379 K. The W specimens were exposed to high-flux (~1 × 1022 m−2 s−1) and high-fluence (~5 × 1025 m−2) D plasma at 873 K in the Tritium Plasma Experiment. Broad desorption peaks extended from 900 K to 1200 K were observed for the neutron-irradiated W by thermal desorption spectroscopy (TDS). The retention in neutron-irradiated specimens was much larger than for an un-irradiated specimen. The highest D retention was obtained for a specimen irradiated at 894 K. With increasing neutron irradiation temperature, the retention was reduced about by half at 1074 K and further increase of the temperature (1379 K) resulted in comparable retention. In addition, one-dimensional diffusion calculations (D desorption in TDS and D depth distribution in plasma exposure) were performed to derive retention parameters (the detrapping energy, the depth occupied by D atoms and D/W ratio) from experimental D retention properties of neutron-irradiated W. By TDS simulation calculation, simple dependences of the peak temperature, height and width of TDS peaks on the retention parameters were obtained with total retention in the orders of 1019 ~ 1022 m−2. The calculation of the depth distribution of trapped D atoms made a relationship between the D/W ratio and the depth occupied by D atoms after plasma exposure at relevant conditions. By comparing the relationship (the D/W and the depth) with that obtained from the experimental results, we estimate each retention parameters for the specimens irradiated by high-dose neutrons at the high temperatures. And, we discussed the neutron-irradiation temperature dependence of D retentions.

QUEST (Q-shu university experiment with steady state spherical tokamak) is a midsize spherical tokamak capable of steady-state operation, comprising all-metal plasma-facing walls and a hot wall (HW) to address issues pertaining to fuel particle balance. The HW was installed summer 2014. Quantitative analysis pertaining to the HW at 373 K is carried out, and clarify the quantitative impact of shot history that obviously appears in wall stored hydrogen just before the discharge at the wall temperature. The model indicates the plasma-induced deposition layer play an essential role in fuel particle balance. A clear temperature dependence of fuel recycling was observed using outgassing just after plasma termination and played an essential role in regulation of particle balance. Consequently, long duration discharges lasting more than 1 h has been obtained at wall temperature, TW

The W (tungsten) samples were placed at top, equator and bottom walls of QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) device and exposed 1238 shots of hydrogen plasma during 2016A/W (Autumn/Winter) campaign with normal wall temperature of 473 K (maximum temperature of 523 K). Thereafter, the surface morphology was evaluated by color measurement, TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). Thick deposition layers were formed on the samples placed at the equator and bottom walls. On the other hand, thin mixed material layer was deposited on the top wall, where large H (hydrogen) retention was observed, which would be caused by dynamic plasma wall interaction (erosion and deposition) with higher H flux. Low H retention was confirmed for bottom wall, where higher wall temperature without He discharge would contribute. The additional 1 keV D2+ was implanted into these samples and deuterium retention enhancement was estimated. It was clearly found that the irradiation damages would induce more deuterium trapping than the formation of C–D bond.

0.8 MeV and 6 MeV iron (Fe) ions were implanted into tungsten (W) to produce the irradiation damages with the various damage distributions. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed to evaluate the D retention behavior on damage distribution in W. The experimental results showed that the total D retentions were decreased by increasing the damage concentration introduced near the surface region by 0.8 MeV Fe ion implantation. The retention of D trapped by vacancy clusters and voids, which would be the stable trapping sites with higher trapping energies, were reduced, suggesting that the recombination of D atom into D2 on the W surface was enhanced due to D accumulation near the surface region. It can be said that the hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface.

Biphasic Li2TiO3–Li4SiO4 tritium breeders with different phase ratios were prepared by the solution combustion synthesis method. The as-prepared biphasic Li2TiO3–Li4SiO4 and single-phase materials were irradiated by the thermal neutron with the flux of 2.75 × 1013 n cm−2 s−1 at Kyoto University Research Reactor (KUR). Tritium is generated based on the following reaction: 6Li+1n→3T+4He+4.8 MeV. The results of tritium release experiment show that tritium release starts at the lower temperature range for the biphasic materials. The tritium release temperature and the shape of tritium-TDS (Thermal desorption spectroscopy) spectra are dependent on the phase ratio of Li2TiO3 to Li4SiO4. In the case of Li2TiO3/Li4SiO4 = 2.0, tritium release starts at 400 K, and the tritium migration is mainly controlled by the diffusion process. With the increase of Li4SiO4 content in the biphasic materials, tritium release is influenced by the de-trapping process.

This research investigated fuel particle balance during long duration discharge in an all-metal plasma facing wall (PFW) through intensive QUEST execution. A simple wall model including the plasma-induced deposition layer that creates hydrogen (H) barriers, called the H barrier model, was established. A simple calculation, based on a combination of H state rate equations and the H barrier model, was applied to real plasma in the early phase of its longest discharge. The model accurately reconstructed the evolutions of electron density and wall-stored H over time, proper values are chosen for the parameters that are difficult to determine experimentally. Comparative calculations that used the H barrier and a fully reflective models, predicted significant impacts of wall models on the plasma density time response and value of electron density, indicating that a proper wall model should be developed for all-metal PFW devices. Keywords: Fuel particle balance, steady state operation, Hydrogen barrier model, QUEST

Tungsten (W) samples were damaged by neutron and 6.4 MeV Fe-ion irradiation above 1000 K simulating the divertor operation temperature. Deuterium (D) retention properties were examined by decorating the damaged W with D and subsequent thermal desorption spectroscopy (TDS) measurements. Vacancy clusters were the major D trapping site in the W irradiated with Fe-ion at 873 K, although D retention by vacancy clusters decreased in the W irradiated with Fe-ion at 1173 K due to dynamic annealing. The D de-trapping activation energy from vacancy clusters was found to be 1.85 eV. D retention in neutron damage W was larger than that damaged by Fe-ion due to the uniform distribution of irradiation defects. The D desorption behaviors from neutron damaged W was simulated well by assuming the D de-trapping activation energy to be 1.52 eV.

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 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.

To estimate the uptake of radiocesium (137Cs) by tea plant roots, 1-year-old rooted tea cuttings (Camellia sinensis L. cv. Yabukita) at the time of bud opening were cultivated hydroponically for 27 days in pots containing nutrient solutions with or without 137CsCl (600 Bq mL−1). Total 137Cs radioactivity of whole tea plants were 6.1 kBq g−1 dry weight. The plant/solution 137Cs transfer factors of different tissues were in the range of 2.6 (in mature leaves) to 28.2 mL g−1 dry weight (in roots), which were lower than those reported in wheat and spinach. In total, 69% of 137Cs remained in roots and 31% was transported from roots to shoots. The results indicated that 137Cs was preferentially translocated to new shoots, which are used for manufacturing tea, over mature leaves.

Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.

To study the hydrogen isotopes plasma driven permeation (PDP) behavior in plasma facing materials, a linear Radio Frequency (RF) plasma device has been constructed in the radiation controlled area at Shizuoka University. The deuterium (D) plasma is generated by injecting RF power with the frequency of 13.56 MHz through a copper antenna and confined by DC magnetic field. The sample is sealed by gold (Au) coated O-ring and one side (upstream side) of sample is exposed to the D plasma. The other side of sample, named as downstream side, is pumped out by a turbo molecular pump and a rotary pump. The permeated D through the sample is monitored by a quadrupole mass spectrometer (QMS) which is connected to the downstream chamber. Infrared heater is adopted to control the sample temperature. The PDP experiments under different plasma parameters show that the permeation process agrees with RD regime. The D recombination coefficient on upstream surface of W is obtained.

Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0 × 1025 m−2 at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9 × 1021 m−2 and near-surface D concentrations up to 1.7 × 10−3 D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0 × 10−3 Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

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.

Due to the low neutron activation and favourable self-passivating properties, Si is introduced to tungsten matrix by spark plasma sintering (SPS) technique. The phase evolution progress and mechanical properties of the composites are investigated. The results show that Si prefers to react with residual oxygen in the powders to form SiOy (y = 1, 1.5, 2) at low Si addition (≤3 at%). In this case, the formation of W5Si3 is suppressed. But when further increasing the Si content up to 32 at%, W5Si3 phase starts to form and its content increases with the elevating ratio of silicon suboxide to SiOy. Combined the results of XPS and tensile testing, it is concluded that silicon oxide is beneficial to the strength of W but not beneficial to the elongation. However, W5Si3 shows positive effects on both strength and elongation. W-32Si composite achieves the best properties with tensile strength 353 MPa and elongation 4.58%. Also, the thermal conductivities were measured, which reveal a growing trend with testing temperature in W-Si composites in contrast with pure W.

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

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.

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

Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He+–3 keV D2+at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He+–1 keV D2+compared to only D2+ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W. Keywords: Simultaneous implantation, D retention, Helium, Flux ratio, Transportation, Thermal desorption spectroscopy

To understand the fuel retention mechanism, correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011-2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load.