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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 D 2 + 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. [ABSTRACT FROM AUTHOR]

• The retention of deuterium (D) for 1 dpa Fe ion damaged W-10%Re with dynamic-annealing or post-annealing was investigated by TDS. • The irradiation defects in W-10%Re by room temperature irradiation made a large D retention. • At higher temperature irradiation, D retention for W-10%Re was reduced less than half for both dynamic-annealing and post-annealing cases. • For post-annealing, higher temperature is required to reduce D retention at 800–1000 K compared to that for dynamic-annealing. The deuterium (D) retention for 1 dpa Fe ion damaged W-10%Re with dynamic-annealing or post-annealing was investigated by TDS. Major D 2 desorption was observed at temperatures of 400, 650 and 850 K for 1 dpa damaged W-10%Re without annealing (room temperature). By dynamic-annealing above 573 K, the D 2 desorption stage at 850 K was clearly eliminated. On the other hand, it was found that post-annealing at a higher temperature of 1173 K was required to annihilate this desorption stage. This indicates that the D desorption from the irradiation defects introduced at room temperature (post-annealing) needs higher annealing temperature compared to the dynamic-annealing. The D retention for dynamic-annealed sample was reduced by the recovery of D trapping sites by annealing. [ABSTRACT FROM AUTHOR]

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 [ABSTRACT FROM AUTHOR]

• Hydrogen (H) and deuterium (D) permeation fluxes for tungsten (W) by H-D mixed plasma were studied to clarify the hydrogen isotope effect on permeation by newly designed plasma driven permeation (PDP) device. • When single species of h or d was irradiated, the h permeation rate showed approximately 1.6 times as high as that of d. • H and d permeation fluxes were clearly reduced by seeding he in H-D mixed plasma. At lower temperature (∼723 K), the ratio of steady permeation fluxes of h and d was about 75 : 25, confirming the high permeation flux of h. • H and d permeation fluxes were reached almost unity at 923 K. Hydrogen (H) and deuterium (D) permeation fluxes for tungsten (W) by H-D mixed plasma were studied to clarify the hydrogen isotope effect on permeation by a newly designed plasma driven permeation (PDP) device at Shizuoka University. A mixed H-D plasma was produced by introducing H and D gasses with various ratios. The atomic ratio and permeation flux in the plasma were measured during the experiments. It was found that the H:D atomic ratio in the plasma was almost proportional to the introduced H:D gas ratio. When single species of H or D was irradiated, the H permeation rate showed approximately 1.6 times as high as that of D. Under H-D mixed plasma irradiation, the highest HD permeation rate was observed when the atomic ratio of H:D in the plasma was 50:50. In addition, the effect of He seeding on H-D mixed plasma driven permeation behavior was also evaluated. The recycling of hydrogen isotopes on the upstream side was enhanced due to the formation of He bubbles near the surface, which would prevent the dissolution of hydrogen isotopes into bulk W and enhance the reflection process during plasma exposure. H and D permeation fluxes were clearly reduced by seeding He in H-D mixed plasma. At lower temperature (∼723 K), the ratio of steady permeation fluxes of H and D was about 75 : 25, confirming the high permeation flux of H. These results have demonstrated a significant hydrogen isotope effect, even if He was seeded or not. As the temperature increased, H and D permeation fluxes approached unity at 923 K. This indicates that the hydrogen isotope effect was determined by the combination of recycling / reflection at or beneath the surface, which will be controlled by the temperature during plasma exposure. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Identifications of tritium trapping states in neutron-irradiated Li 1.8 TiO 2.9 (lithium-depleted Li 2 TiO 3 ) were carried out by the out-of-pile tritium release behavior. Tritium release behaviors for neutron-irradiated Li 2 TiO 3 and tritium gas-exposed TiO 2 were also measured for comparison. Among the tritium release spectra for these samples, three tritium release peaks were appeared. By the kinetic analyses of tritium release behaviors, the Arrhenius parameters for three peaks were evaluated. Especially for Li 1.8 TiO 2.9 , there were two tritium release peaks, and the peak in lower temperature region was assigned to the tritium release controlled by the diffusion process in Li 2 TiO 3 structure. The other tritium release peak, which was hardly appeared for Li 2 TiO 3 , was assigned to the release of tritium trapped as hydroxyl groups in Li 1.8 TiO 2.9 , indicating that lithium-depletion would result in the formation of hydroxyl groups in Li 2 TiO 3 . Lithium vacancies existed in Li 2 TiO 3 crystal structure would promote the tritium trapping as hydroxyl groups due to the decreased charge repulsion between lithium ions and tritium ion, resulting in the difficulty of recovering tritium from Li 2 TiO 3 effectively. [ABSTRACT FROM AUTHOR]

This paper reviews recent results pertaining to hydrogen isotope behavior in neutron and heavy ion damaged W. Accumulation of damage in W creates stable trapping sites for hydrogen isotopes, thereby changing the observed desorption behavior. In particular, the desorption temperature shifts higher as the defect concentration increases. In addition, the distribution of defects throughout the sample also changes the shape of TDS spectrum. Even if low energy traps were distributed in the bulk region, the D diffusion toward the surface requires additional time for trapping/detrapping during surface-to-bulk transport, contributing to a shift of desorption peaks toward higher temperatures. It can be said that both of distribution of damage (e.g. hydrogen isotope trapping sites) and their stabilities would have a large impact on desorption. In addition, transmutation effects should be also considered for an actual fusion environment. Experimental results show that production of Re by nuclear reaction of W with neutrons reduces the density of trapping sites, though no remarkable retention enhancement is observed. [ABSTRACT FROM AUTHOR]

After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handling capacity in facilities or universities, present activities on T engineering research in Japan are summarized in short in terms of T accountability on safety. The term of safety includes wide processes from T production, assay, storing, confinement, transfer through safety handling finally to shipment of its waste. In order to achieve reliable operation of fusion reactors, several unit processes included in the T cycle of fusion reactors are investigated. Especially, the following recent advances are focused: T retention in plasma facing materials, emergency detritiation system including fire case, T leak through metal tube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement are especially demanded for T accountability depending on various molecular species. Since kg-order T of vaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accuracy of <0.1% is demanded far over the conventional technology status. Necessity to control T balance within legal restrictions is always kept in mind for operation of the future reactor. [ABSTRACT FROM AUTHOR]

W samples were irradiated at 300 K with 6 MeV Fe ion with damage concentrations in the range from 0.0003 to 1.0 displacements per atom (dpa) and then implanted at 300 K with 500 eV D ions to a fluence of 5 × 10 21 D/m 2 . Deuterium retention in the damaged samples was examined in situ by thermal desorption spectrometry (TDS). Simulation of the TDS spectra was performed using the Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code to reveal the binding energies for deuterium captured by the ion-induced defects. It has been shown that the deuterium TDS spectra consist of two or three peaks (depending on the damage concentration) at about 400, 600 and 800 K, and can be simulated by the HIDT simulation code with the use of hydrogen-trap binding energies of 0.65, 1.25, and 1.55 eV. [ABSTRACT FROM AUTHOR]

The annihilation kinetics of irradiation-induced defects in Li 2 TiO 3 was evaluated by electron spin resonance. The radiation defects were stabilized with increasing defect density by interaction with neighboring defects. Subsequently, the tritium migration model in Li 2 TiO 3 was established by integrating the kinetics of tritium diffusivity in Li 2 TiO 3 crystalline grains, tritium trapping/detrapping at oxygen vacancies and hydroxyl groups, and annihilation of irradiation-induced defects. The contribution of hydrogen isotopes contained in tritium recovery gas was also considered in this model. The model can demonstrate an overall profile of out-of-pile tritium release for Li 2 TiO 3 with various neutron fluences. Tritium release behavior under neutron irradiation was also estimated by the present established model. The tritium inventory increased under high neutron flux because of the continuous generation of tritium trapping sites, and the addition of hydrogen isotopes to the purge gas could reduce the tritium inventory for Li 2 TiO 3 . [ABSTRACT FROM AUTHOR]

The green tea leaf was dipped in the CsCl solution to elucidate the dissolution behavior of Cs. The IP observation showed that Cs was not uniformly adsorbed on the tea leaf, even if the tea leaf was dipped in the homogeneous solution. The activation energy of Cs dissolution was estimated to be 0.045 eV, indicating that most of Cs would exist as the adsorbed state. The dissolution behaviors as a function of cesium and potassium concentrations and pH, were also studied. [ABSTRACT FROM AUTHOR]

Highlights: [•] The carbon-dominant mixed-material layer was formed on the tungsten surface after 15th plasma exposure in LHD. [•] The largest enhancement of deuterium retention was found to be about 21. [•] The major deuterium desorption temperature was shifted toward higher temperature side. [Copyright &y& Elsevier]

Abstract: Migration mechanisms of hydrogen isotopes in Li2TiO3 were investigated by thermal desorption spectroscopy. The apparent tritium release kinetics of thermal-neutron-irradiated Li2TiO3 was mainly governed by tritium diffusion in Li2TiO3 crystal grains. The irradiation defects and O–D bonds induced in Li2TiO3 by energetic deuterium ion implantation were also investigated by X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy, respectively. The deuterium release spectra for Li2TiO3 irradiated with energetic deuterium ions consisted of five release peaks. The rate-determining processes for these release peaks were not only deuterium diffusion but also the detrapping of deuterium in irradiation defects and in the dissociation of O–D bonds (hydroxyl groups). The detrapping was the dominant deuterium release mechanism for energetic deuterium ion irradiation because of the shorter diffusion length. The rates of diffusion, detrapping from irradiation defects, and detrapping from hydroxyl groups were determined, and the rate-determining process of hydrogen isotope release in Li2TiO3 crystalline grains of various sizes was assessed. [Copyright &y& Elsevier]

Abstract: Hydrogen isotope retention and desorption behaviors for Silicon carbide (SiC), Silicon nitride (Si3N4) and Silicon dioxide (SiO2) were studied to elucidate the fundamental process of hydrogen isotope in Si related ceramics by means of T-IP (tritium imaging plate), thermal desorption spectroscopy (TDS) and X-ray photoelectron spectroscopy (XPS). The tritium gas exposure at 673K showed that tritium was precipitated on the surface for SiO2, although that for SiC was uniformly retained inside the bulk. The 0.2keV implantation revealed that the deuterium desorption stages for Si related ceramics were consisted of four desorption stages at around 450K, 650K, 800K, and 950K, attributing to the desorptions of deuterium trapped on the surface, retained in interstitial sites, trapped as Si–D bond and trapped as C/N/O–D bond, respectively. The retention enhancement of deuterium trapped by Si as Si–D bond and the reduction of deuterium trapped on the surface would be associated with the enhancement of covalent bond characteristics for Si related ceramics. These results indicate that the dangling bonds in the covalent ceramics have higher hydrogen isotope trapping efficiency to form chemical bond like Si–D bond. On the other hand, the surface adsorption of hydrogen isotope was enhanced for the higher ionicity ceramics by charge localization. [Copyright &y& Elsevier]

Abstract: Effect of surface oxide layer on the hydrogen isotope permeation was studied. Iron oxide was uniformly formed in the oxide layer, although chromium was limited at the interface between the oxide layer and bulk SS-316. The permeation behavior of deuterium for oxidized SS-316 was compared with that for unoxidized SS-316 at temperature range of 333–673K. The deuterium permeability for the oxidized SS-316 was reduced 1/10–1/20 times as high as that for unoxidized one. However, the activation energy of deuterium permeation as gas form for oxidized SS-316 was almost the same as that for unoxidized SS-316 and was 0.64eV, which was almost consistent with the sum of activation energies for diffusion and solubility. This fact indicates that the deuterium permeation is diffusion limited. The permeability of deuterium as water form was almost constant even if heating temperature is high, showing that the deuterium was permeated through bulk SS-316 and react with oxygen at the oxide layer as water desorption, which is controlled by the permeation flux of deuterium and oxygen concentration on the surface of oxide layer in downstream side. [Copyright &y& Elsevier]

Abstract: Behaviors of hydrogen isotope retention and damages in tungsten and SS-316 with simultaneous C+–D2 + implantation were compared to those with only D2 + implantation using X-ray photoelectron spectroscopy (XPS), Thermal desorption spectroscopy (TDS), glow discharge-optical emission spectroscopy (GD-OES) and Transmission electron microscopy (TEM). The total D retention for SS-316 with only D2 + implantation was about 45% as large as that for tungsten. The D retention for simultaneous C+–D2 + implanted tungsten and SS-316 clearly increased as a factor of 1.7, which is almost the same among these samples. The density of dislocation loops was enhanced by the simultaneous C+–D2 + implantation, indicating the D trapping site would be produced by C+ implantation. As for the D desorption temperature, small shift toward lower temperature side was found for SS-316 compared to tungsten, indicating the D trapping energy by dislocation loops and grain boundary for SS-316 is lower than that for tungsten. [Copyright &y& Elsevier]

Abstract: The ion fluence dependence of deuterium retention in SS-316 during oxidation at a temperature of 673K was studied to evaluate the dynamics of deuterium retention in the oxide layer of SS-316. The correlation between the chemical state of stainless steel and deuterium retention was evaluated using XPS and TDS. It was found that the major deuterium desorption temperatures were located at around 660K and 935K, which correspond to the desorption of deuterium trapped as hydroxide. The deuterium retention increased with increasing deuterium ion fluence, since the deuterium retention as hydroxide increased significantly. However, retention saturated at an ion fluence of ∼2.5×1021 D+ m−2. The XPS result showed that FeOOD was formed on the surface, although pure Fe also remained in the oxide layer. These facts indicate the nature of the oxide layer have a key role in deuterium trapping behavior. [Copyright &y& Elsevier]

Abstract: To elucidate the C+ implantation effect on deuterium retention in tungsten, simultaneous C+ and implantation was performed on tungsten. Deuterium desorption behavior for simultaneously implanted tungsten was compared with that the sequentially implanted tungsten. It was found that deuterium retention for the simultaneous case was lower than that for the sequential one, indicating that the sputtering of carbon and deuterium trapped by carbon would prevent the deuterium trapping. In addition, deuterium retention decreased with increasing C+/D+ ratio and deuterium trapping by carbon was not observed for high C+/D+ ratio. These facts indicate that the number of available traps in tungsten produced by simultaneous ion implantation would mainly control the deuterium retention in tungsten. However, for low C+/D+ ratios, deuterium trapping by carbon would be one of major trapping states. [Copyright &y& Elsevier]

Abstract: The implantation temperature dependence of hydrogen isotope behaviors in non-helium ion (He+) pre-implanted silicon carbide (SiC) and He+ pre-implanted SiC was studied by means of X-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS). It was found that the deuterium (D) retention decreases as the implantation temperature increases in both non-He+ pre-implanted SiC and He+ pre-implanted SiC after D2 + implantation. Above 800K, D bound only to C in forming C–D bonds. The chemical states of Si 2p for non-He+ pre-implanted SiC and He+ pre-implanted SiC at implantation temperatures above 800K are quite different from those at room temperature and a positive peak shift of Si 2p was observed, indicating Si–C–D bonds were formed without larger amount of retention of damaged structures. These facts indicate that damaged structures introduced by D2 + and/or He+ implantation are quickly recovered by the thermal annealing effect at temperature above 800K. These results imply that the thermal annealing effect influence widely on the tritium behavior and damaged structures in SiC. [Copyright &y& Elsevier]

Abstract: Four Buckingham-type pairwise potential models for Li2O were assessed by molecular static and dynamics simulations. In the static simulation, all models afforded acceptable agreement with experimental values and ab initio calculation results for the crystalline properties. Moreover, the superionic phase transition was realized in the dynamics simulation. However, the Li diffusivity and the lattice expansion were not adequately reproduced at the same time by any model. When using these models in future radiation simulation, these features should be taken into account, in order to reduce the model dependency of the results. [Copyright &y& Elsevier]

Abstract: To elucidate the interaction between hydrogen isotopes and damaged structures in SiC, deuterium ion was implanted into helium ion (He+) implanted SiC. The chemical state of SiC was analyzed by XPS and thermal desorption behaviors of D2 and He were observed by TDS. It was found that D desorption consisted of two stages, namely D bound to Si and that to C. The D was desorbed from both desorption stages for He+ implanted SiC with the low fluence, although it was preferentially trapped by C for pure SiC. The D retention for He+ implanted SiC was lower than that for pure SiC under the same fluence. The implanted He mainly forms He blister or remains in the carbon vacancies. The D trapping efficiency was changed by He+ implantation, although the D trapping mechanism was not largely influenced. [Copyright &y& Elsevier]

Abstract: Ultraviolet/X-ray photoelectron spectroscopy was performed for a polycrystalline pellet sample of Li2O, in order to elucidate the interaction of H2O with the Li2O surface. By H2O-gas exposures of 0–13L (1L=1.33×10−4 Pas), a variation was induced in the electronic structure, corresponding to the formation of surface –OH via the dissociative adsorption of H2O. No clear qualitative difference arose either in the electronic structure or in the reactivity to H2O between the polycrystalline pellet and the Li2O (111) single crystal. The variation in the contribution of –OH to the total intensity of the O 1s peak was investigated as a function of the amount of H2O-gas exposure, and the result was discussed in conjunction with an adsorption/absorption model using a Monte Carlo simulation. It was indicated that not only adsorption, but also absorption up to a few surface layers occurred with a high probability even in the annealed sample. [Copyright &y& Elsevier]

Abstract: In order to understand the influence of radiation defects on hydrogen isotope behavior in Li2O, thermal desorption spectroscopy (TDS) was applied to single-crystal samples into which deuterium had been introduced by two different methods: deuterium ion irradiation or deuterium thermal absorption. In both cases, a sharp peak of water species was observed at 680K due to the decomposition of the LiOD phase. Additionally, two desorption peaks, one corresponding to HDO at 870K and the other to HD/D2 at 900K, were found for the irradiated sample. A comparison with the literature regarding these two desorption peaks at this stage suggested that the former desorption was due to D+ diffusing in the bulk, while the latter was due to D− trapped by the F0 centers or the deuterium bubbles. [Copyright &y& Elsevier]

Abstract: In order to understand the influence of radiation defects on hydrogen isotope behavior in Li2O, an in situ IR absorption analysis was carried out during 3keV irradiation. Three IR absorption peaks were obtained from the stretching vibrations of O–Ds in different existence states: an O–D existing as a LiOD phase at 2710cm−1, a surface O–D with several adjacent O–Ds at 2660cm−1, and an O–D affected by a defect in the bulk at 2605cm−1. At the outset of the irradiation, only the surface O–D appeared. As the ion fluence increased, the O–D affected by a defect started to form in the bulk, and the LiOD phase followed it. The defect related with the O–D at 2605cm−1 was regarded as a Li vacancy. It was indicated that the hydrogen isotopes were stabilized in Li2O through an interaction with the Li vacancy (LiO−–D+: 2605cm−1) or mutual aggregation (LiOD phase: 2710cm−1). [Copyright &y& Elsevier]

In order to estimate the stabilities of various O–D states formed in Li2O by 3 keV D2+ irradiation, heating experiments up to 453 K were conducted. The chemical states of O–D bonds were analyzed by FT-IR, and the variation of the intensities of the O–D bonds in various chemical states were studied individually. During the irradiation, two O–D states were mainly formed in the bulk: O–D bond existing as LiOD phase or that without forming the LiOD phase. As sample temperature increased, transition of the latter to the former was observed. The apparent activation energy for the decrease of the O–D bond without forming the LiOD phase was estimated to be 60 kJ/mol over the temperature range 320–362 K, and that existing as the LiOD phase was estimated to be 130 kJ/mol over 403–453 K. [Copyright &y& Elsevier]

The retention and isotope replacement behavior of hydrogen isotopes in SiC has been studied by means of the elastic recoil detection (ERD) technique. The protium ions or deuterium ions were implanted into SiC at room temperature up to almost the saturation. Successively, the ion source was switched to deuterium ions or protium ions and the samples were irradiated at the same fluence under the same experimental conditions, respectively. The mass balance equations were applied to analyze the retention curves and the curves of decay and uptake in the isotope replacement. The saturation concentrations of H and D have been found to be about 0.70 and 0.75, respectively. The detrapping cross sections of H and D implants by the D2+ and H2+ ion bombardments have been determined to be 3.2±0.3×10−22 m2/D+ and 2.6±0.2×10−22 m2/H+, respectively. The magnitude of these detrapping cross sections from SiC is as large as that from graphite. These isotope differences are discussed in terms of the elementary processes included in the mass balance equations. [Copyright &y& Elsevier]

• Hydrogen isotope exchange behavior on C-W mixed material was studied by TDS. • The temperature dependence showed that hydrogen isotope exchange was the most effective at 523 K. • It can be said that hydrogen isotope replacement is efficient to remove retained tritium on the surface region of mixture layer. Hydrogen isotope exchange behavior on carbon (C)- tungsten (W) mixed material layer on W was studied as a function of gas exposure time and temperature. It was found that hydrogen isotope exchange enhanced the reduction of residual hydrogen effectively compared by only heating. The temperature dependence showed that hydrogen isotope exchange was the most effective at 523 K and above 573 K, thermal annealing reduced the total hydrogen isotope retention. It can be said that hydrogen isotope replacement is efficient to remove retained tritium on the surface region of mixture layer. [ABSTRACT FROM AUTHOR]

At present, the biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramic pebbles with high lithium content and crushing load have been viewed as promising tritium breeders for maintaining the safety-state operation of the D-T fusion reactor. The tritium release behavior of Li 2 TiO 3 –Li 4 SiO 4 ceramic pebbles fabricated by centrifugal granulation method was investigated by using the Thermal Desorption Spectroscopy apparatus at Shizuoka University. It can be found that the Li 2 TiO 3 –Li 4 SiO 4 ceramic pebbles show three tritium release peaks, and most of the tritium was released as HTO observed from T-TDS spectra, and the peaks at 581 K and 734 K are mainly due to tritium release from Li 2 TiO 3 in core and shell. The peak at 813 K is mainly deemed as tritium release derived from the Li 4 SiO 4 shell. The content of tritium release from biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramic is up to 11.5 MBq g−1, but the amount of gas species (HT) is only 0.126 MBq g−1. Besides, the isothermal heating experiment showed that the tritium release behavior was subject to diffusion process and de-trapping in defect sites. At the same time, the irradiation defects of E′-center, O-related center, and Ti3+ for breeding material can be found by using an ESR device, and the isochronous heating experiments have manifested that the amount of irradiated defect decreases as heating temperature increases, which indicates that the tritium release behavior is closely related to defect annihilation process. Therefore, the advanced Li 2 TiO 3 –Li 4 SiO 4 ceramic breeder can be fabricated based on tritium release result, which will be conducive to meet the practical application of breeder in the future. [ABSTRACT FROM AUTHOR]

Lithium ceramics are regarded as potential candidates for tritium breeding in fusion reactors. Lead (Pb) was added to the ceramic to improve the strength and tritium release properties. The Li 2 TiO 3 -0.5Li 4 SiO 4 –Pb ceramic powders with the Pb contents of 5 wt% and 20 wt% were prepared by microwave-induced solution combustion synthesis. It was confirmed that the Pb was in the form of the Li 2 PbO 3 phase in the powders. The Li 2 TiO 3 , Li 4 SiO 4 , and Li 2 PbO 3 crystallites were composited at nanoscale level, which was confirmed by the transmission electron microscopy (TEM). During high-temperature sintering, the Li 2 PbO 3 was transformed into PbO, which facilitated the densification process at low temperatures. The chemical state in the sintered body was characterized by X-ray photoelectron spectroscopy (XPS) with 3 keV Ar+ sputtering. The binding energies for Pb were 137.2 and 142.2 eV. There was almost no change in peak position during Ar+ sputtering, showing that Pb was firmly adherent with oxygen atoms. Besides, the as-synthesized powders were used for the pebble fabrication. The average crush load of Li 2 TiO 3 -0.5Li 4 SiO 4 –20%Pb pebbles sintered at 1223 K was 124 N, which was much higher than that without Pb addition. It was proved that the crush load increased almost linearly with the increase in the size of the pebbles. Thus, the bending strength of the sintered ceramic body was used to evaluate the mechanical properties. Further, the grain growth kinetics was studied. It was confirmed that the Pb addition reduced the activation energy of grain growth, which realized the densification of the ceramic pebbles at low temperatures. [ABSTRACT FROM AUTHOR]

The neutron multiplier lead (Pb) containing Li 2 TiO 3 -0.5Li 4 SiO 4 –Pb was prepared, which aimed at the improvement of the tritium release performance of the lithium ceramic without an increase in lithium density. Li 2 TiO 3 -0.5Li 4 SiO 4 –Pb and that without Pb were irradiated at Kyoto University Research Reactor (KUR) with the thermal neutron flux of 5.5 × 1012 n s−1 cm−2 and the fast neutron flux of 1.2 × 1012 n s−1 cm−2. Fast neutrons can trigger the Pb (n , 2 n) reaction. The tritium release experiments were conducted using the tritium thermal desorption spectroscopy (tritium-TDS) system. Compared with Li 2 TiO 3 -0.5Li 4 SiO 4 , the tritium release rate at the lower temperature range was higher than that for Li 2 TiO 3 -0.5Li 4 SiO 4 –Pb. The tritium-TDS spectrum was also shifted to the lower temperature side. The isothermal heating experimental results confirmed that tritium in the form of gas species (HT) was controlled by the diffusion process, while the trapping and de-trapping process contributed to the total tritium release. Further, 3 keV D 2 + implantation with the fluence of 2.0 × 1022 D m−2 and subsequent TDS measurements were performed. It was shown that the D 2 release at the lower temperature side was increased for Li 2 TiO 3 -0.5Li 4 SiO 4 –Pb. Meanwhile, HD and HDO release at higher temperature side was significantly decreased. [ABSTRACT FROM AUTHOR]

The tritium breeder and structural materials are necessary components in the blanket to realize tritium (T) self‐sustainment of nuclear fusion. The long‐term exposure between tritium breeders and structural materials will cause surface corrosion in irradiation environments and then further affect the tritium release behavior. In this study, chemical compatibility between Li2TiO3 ceramic pebbles and advanced structural materials was studied systematically at 700 °C for 300 h under He+0.1 % H2 environment, respectively. The color of the Li2TiO3 ceramic pebbles changes from white to dark grey and black. Moreover, the grain size of Li2TiO3 ceramic pebbles increases to more than 5 μm, and the crushing load decreased slightly. For the structural materials, the Al‐rich oxide layer with about 188.7 nm of 14Cr‐5Al oxide dispersion strengthened (ODS) steel and Cr‐Fe rich oxide layer with about 1.04 μm of 14Cr‐ODS steel were observed on the cross‐section. The effective diffusion coefficient of O element in Li2TiO3 ceramic moved into ODS steel at 700 °C was calculated to be 3.3 × 10−16cm2/s and 1.02 × 10−14 cm2/s. Unfortunately, when SiC ceramics were contacted with the pebbles, the crystal phase transformed into SiO2, which severely limits its application. Therefore, these results will provide guidance for the selection of structural materials in the future. [ABSTRACT FROM AUTHOR]

Abstract: The deuterium (D) retention for Vacuum Plasma Spray (VPS)–tungsten (W)/F82H was studied using two different implantation methods, namely D plasma exposure and implantation. The D retention for polished VPS–W/F82H after plasma exposure was found to be reduced compared to that for polycrystalline tungsten. Scanning electron microscopy (SEM) and transmission electron microscopy (TEM) observations indicated that porous structures around grain boundaries and the interface between VPS–W layers would be potential D diffusion paths, leading to low D retention. In the case of implantation, the shape of D2 TDS spectrum was almost the same as that for D plasma-exposed VPS–W/F82H; however, the D retention was quite high for unpolished VPS–W/F82H, indicating that most of D was trapped by the oxide layer, which was produced by the VPS process. The reduction of surface area due to the polishing process also reduces D retention for VPS–W/F82H. These results indicate that controlling the surface chemical states is important for the reduction of tritium retention for future fusion reactors. [Copyright &y& Elsevier]

Abstract: Tungsten (W) samples, undamaged and damaged by irradiation with 2.4MeV Cu2+, were placed flush to the first wall of Large Helical Device (LHD) nearby the divertor target, and were exposed to 134 shots of hydrogen plasma discharges. Thereafter, to evaluate the enhancement of hydrogen isotope retention for W due to LHD hydrogen (H) plasma exposures, the W samples with and without LHD H plasma exposures were implanted with 1.0keV . For W samples without LHD H plasma exposures, the D retention in the damaged W was by a factor of ∼6 higher than that in the undamaged W. After LHD H plasma exposures, the carbon layers ∼4nm in thickness were deposited on the W surfaces. Due to trapping of D atoms in the carbon layers, the D retention enhancement for undamaged and damaged W samples exposed to LHD H plasmas was significantly higher than that for the W samples without LHD H plasma exposures. [Copyright &y& Elsevier]

Hydrogen isotope migration behaviors in ceramics-metal multilayer coatings have been elucidated for a further improvement of fueling system and radiological safety. Erbium oxide (Er 2 O 3 ) 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 Er 2 O 3 coating. An Er 2 O 3 -Fe-Er 2 O 3 three-layer coating was also fabricated by the VAD on the Er 2 O 3 -Fe coating. The grain boundary diffusion model was constructed based on the results of grain structure analysis and deuterium depth profile for the Er 2 O 3 single-layer coating from our previous works. The Er 2 O 3 -Fe two-layer coatings with different Fe layer thickness showed no significant difference in deuterium permeability. The Er 2 O 3 -Fe-Er 2 O 3 three-layer coating showed a PRF of up to 10 4 due to a contribution of two diffusion barriers of the inner and outer Er 2 O 3 layers. In addition, the three-layer coating had three times higher deuterium concentration than the Er 2 O 3 -Fe coating, although the concentration will be negligibly small from the perspective of tritium inventory in the fuel system. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Abstract: To elucidate the simultaneous implantation effects on deuterium retention in polycrystalline tungsten, simultaneous 10keV C+, 3keV and 3keV He+ implantation was performed at room temperature. He+/D+ flux ratio dependence of deuterium retention indicated that He+ implantation induced high deuterium retention even if the He+/D+ flux ratio was low. D2 TDS spectrum for C+– implanted tungsten was clearly different from the samples with only , –He+ and C+––He+ implantation, indicating that D was trapped by C as C–D bond. For –He+ implantation sample, He+ implantation would introduce irradiation damages and D retention increased. However, C+––He+ implantation, the accumulation of C on tungsten was suppressed and the retention of D trapped by C clearly decreased. These facts indicated the D retention for C+––He+ implanted tungsten was limited by He+ implantation and most of D would be retained in the interstitial sites or irradiation defects under simultaneous implantation with He. [Copyright &y& Elsevier]

Neutron irradiation with different fluence and flux on the Li 2 TiO 3 -0.5Li 4 SiO 4 biphasic ceramic powders was conducted. The tritium release behavior was studied by thermal desorption spectroscopy (TDS) with different heating rates and isothermal heating at various temperatures. It was found that the tritium TDS spectra were shifted toward the higher temperature side as the neutron fluence was reduced. The tritium migration in the biphasic ceramic was governed by diffusion and de-trapping processes with the activation energies of 0.29–0.50 eV and 1.00–1.13 eV for the neutron-irradiated samples with the fluence of 8.25 × 1015 and 3.96 × 1016 n cm−2. For isothermal heating, there is another tritium desorption peak when the sample was continually heated to higher temperatures after isothermally heating at 690 K and 800 K, indicating that tritium still remained in the ceramic under isothermal heating conditions. The intensity of the second peak was decreased at the higher isothermal heating temperatures. The diffusion model with detrapping reaction can fit the experimental data, suggesting that both the detrapping and diffusion processes should be considered to estimate tritium release characteristics for the tritium breeding biphasic ceramics. • Neutron irradiation on the biphasic ceramic was performed. • Tritium desorption temperature range became narrow at high neutron flux. • Tritium release peak temperature increased at higher neutron fluence. • Tritium release from biphasic ceramic was affected by detrapping reactions. [ABSTRACT FROM AUTHOR]

Lithium ternary ceramics are regarded as promising candidates for tritium breeding in the fusion reactor. In this study, the neutron multiplier element, lead (Pb), was mixed into Li 2 TiO 3 -Li 4 SiO 4 ceramics to enhance their tritium breeding performance and mechanical properties. The Li 2 TiO 3 -Li 4 SiO 4 -Pb ceramic powders were synthesized via microwave-induced solution combustion, employing citric acid as the fuel to composite elemental Pb into the ceramics. The ceramic powders were calcined under pure oxygen for the removal of the Li 2 CO 3 impurity and the improvement of the crystallization. The characterization results show that elemental Pb is homogeneously distributed throughout the ceramic body mainly in the chemical form of Li 2 PbO 3. Compared to the conventional Li 2 TiO 3 -Li 4 SiO 4 biphasic ceramics, the Li 2 TiO 3 -Li 4 SiO 4 -Pb ceramics sintered at 1173 K exhibits higher bending strength. The exposure to argon purging gas at high temperatures was conducted to examine the phase stability and mechanical performance with the consideration of tritium recovery working conditions. There is no significant change in the phase composition and bending strength with the exposure duration of up to 60 h. [ABSTRACT FROM AUTHOR]

Abstract: Tritium migration behaviors in Li2TiO3 with the increase of irradiation damage density were investigated by means of electron spin resonance and thermal desorption spectroscopy. The irradiation damages of F+-centers and O−-centers were formed by neutron irradiation, and their damage densities were increased with increasing neutron fluence. Tritium release temperature was clearly shifted toward higher temperature side with increasing neutron fluence, i.e. increasing damage density. The rate determining process for tritium release was also clearly changed depending on the damage density. Tritium release was mainly controlled by tritium diffusion process in crystalline grain of Li2TiO3 at lower neutron fluence. The apparent tritium diffusivity was reduced as the damage density in Li2TiO3 increased due to the introduction of tritium trapping/detrapping sites for diffusing tritium. Then, tritium trapping/detrapping processes began to control the overall tritium release with further damage introductions as the amount of tritium trapping sites increased enough to trap most of tritium in Li2TiO3. The effects of water vapor in purge gas on tritium release behaviors were also investigated. It was considered that hydrogen isotopes in purge gas would be dissociated and adsorbed on the surface of Li2TiO3. Then, hydrogen isotopes diffused inward Li2TiO3 would occupy the tritium trapping sites before diffusing tritium reaches to these sites, promoting apparent tritium diffusion consequently. Kinetics analysis of tritium release for highly damaged Li2TiO3 showed that the rate determining process of tritium release was the detrapping process of tritium formed as hydroxyl groups. The rate of tritium detrapping as hydroxyl groups was determined by the kinetic analysis, and was comparable to tritium release kinetics for Li2O, LiOH and Li4TiO4. The dangling oxygen atoms (O−-centers) formed by neutron irradiation would contribute strongly on the formation of hydroxyl groups. The efficiency of tritium trapping/detrapping by the dangling oxygen atoms was clearly increased with increasing damage density due to the stabilization of damages by neighboring irradiation damages and/or the lithium burn-up which produces lithium vacancy acting as a pass way of tritium to the dangling oxygen atoms. [Copyright &y& Elsevier]

• Neutron irradiation and tritium release experiments were performed. • A tritium release model for the biphasic ceramic was validated. • The kinetic parameters are consistent with the experiment data. • In-pile tritium release data can be obtained using the model. • Controlling the temperature of the blanket system can reduce tritium retention. In this study, neutron irradiation for the Li 2 TiO 3 -Li 4 SiO 4 biphasic ceramics was conducted at Kyoto University Research Reactor (KUR). The tritium thermal desorption spectroscopy (tritium-TDS) and the electron spin resonance (ESR) were performed in the radiation-controlled area at Shizuoka University. It was confirmed that the tritium desorption peculiarities were dependent on the phase ratio of Li 2 TiO 3 and Li 4 SiO 4 in the ceramic and the neutron fluence. The tritium migration process in biphasic ceramics was intensively studied based on the diffusion and de-trapping kinetics. A tritium release model was established with the consideration of the tritium trapping by irradiation defects and diffusion/migration process from bulk to surface. The experimental data can be well reproduced by the proposed simulation code. The kinetic parameters obtained by the simulating process were almost consistent with the experimental results. The effects of grain size and phase ratio on the tritium release characteristics were also investigated. The simulation code can be also applied to pellets or pebbles. Besides, the simulation code was adapted to understand the tritium release behavior in the working condition of fusion reactors. It was found that the tritium release could reach a steady state when the blanket was operated at higher temperatures. The tritium retention in ceramics after the shutdown of the reactor was confirmed. The tritium inventory can be reduced by controlling the temperature of the blanket system. [ABSTRACT FROM AUTHOR]

The effect of Pb addition on tritium recovery for Li 2 TiO 3 -Li 4 SiO 4 mixed ceramic materials was investigated by Tritium-thermal desorption spectroscopy (T-TDS). The behavior of irradiation defects was also studied by electron spin resonance (ESR) with heating. The sample with the lower neutron fluence of 8.0 × 1015 n cm−2 had a tritium desorption stage on the low temperature side, 400 K, which can be ascribed to the tritium adsorption on the surface. As the neutron fluence was increased, the desorption of tritium retained in the bulk was major tritium desorption process. The major desorption stage was shifted to the lower temperature side by Pb addition. The intensity of ESR peak corresponding to O −-center was reduced by Pb addition, indicating the irradiation defect was decreased. This contributes on the reduction of tritium trapping sites and facilitates tritium recovery on the lower temperature side. [ABSTRACT FROM AUTHOR]

Abstract: The behavior and existence states of deuterium irradiated into LiNbO3 and LiTaO3 single crystals were studied by in situ FT-IR observations on multiple O–D vibration peaks and by TDS on the desorption behavior of deuterium. Two O–D vibration peaks were observed in both LiNbO3 and LiTaO3, one at 2610–2615cm−1 and the other at 2560–2570cm−1. The former O–D vibration was attributed to –OD− affected by Nb defect in LiNbO3 or Ta defect in LiTaO3. For attribution of the latter O–D vibration, two candidates remained, –OD− affected by Li vacancy and –OD− of interstitial D+. By heating after the ion irradiation, the irradiated deuterium was mainly desorbed as hydrogen molecular forms (D2 and HD). The –OD− affected by Nb defect in LiNbO3 or Ta defect in LiTaO3 had a higher desorption temperature than the other –OD−s. The two materials showed similar results in the existence states and the desorption behavior of irradiated D+, although some differences were observed due to more formation of Nb defect in LiNbO3 than Ta defect in LiTaO3 by displacement. It was indicated that the release temperature of hydrogen isotopes is heightened by the influence of M defect in ternary lithium oxides (Li x MO y ). [Copyright &y& Elsevier]

Abstract: The existence states of deuterium in LiAlO2 were analyzed by in situ IR absorption spectroscopy during irradiation with 3keV at room temperature. Multiple IR absorption peaks that were related to O–D stretching vibrations were observed, mainly at 2650cm−1 (O–Dα), 2600cm−1 (O–Dβ), and 2500cm−1 (O–Dγ). The O–Dα was assigned to the surface O–D. The O–Dβ and O–Dγ were interpreted as two distinct O–D states for three candidates: O–D of substitutional D+ for Li+; O–D of substitutional D+ for Al3+; and O–D of interstitial D+. O–Dβ was the dominant O–D state for deuterium irradiated into LiAlO2, and had higher stability than O–Dγ. Heating after ion irradiation led to the desorption of D2 and an increase in the intensity of O–Dβ, which implies that some of the deuterium irradiated into LiAlO2 exists in non-O–D states, such as D− captured by F centers. [Copyright &y& Elsevier]

Abstract: Surface variation of Li2TiO3 induced by Ar+ sputtering, heating, or water vapor exposure was analyzed using X-ray/ultraviolet photoelectron spectroscopy (XPS/UPS). For a near-stoichiometric oxide composition, such as that produced by Ar+ sputtering and annealing at 673K, little or no adsorption of hydroxyl groups (−OH) occurred by 5×10−3 Pa H2O exposure. Conversely, dissociative (−OH) adsorption was observed for reduced surface created by Ar+ sputtering via selective sputtering of Li and O. The surface nature of Li2TiO3 showed analogous behavior with that of TiO2 rather than Li2O. [Copyright &y& Elsevier]

The hydrogen isotope behavior in Li2O at low temperature was studied by Fourier transform infrared spectroscopy (FT-IR) to clarify the fundamental behavior of hydrogen isotopes. In the IR spectra obtained under the deuterium ion irradiation at 228 K, three peaks at 2710, 2660 and 2605 cm-1, and a broad peak at a region of 2700–2500 cm-1 were observed, corresponding to the O–D stretching vibrations. The O–D bonds as the broad peak changed to the LiOD phase as the temperature increased. In comparison with the results of irradiation at room temperature, (i) the formation of LiOD phase was restrained, and (ii) the broad peak could be apparently observed only in the low temperature irradiation. Therefore, it was concluded that the chemical states of hydrogen isotopes in Li2O have a dependence on the sample temperature of the deuterium ion irradiation. [Copyright &y& Elsevier]

Tungsten (W) is one of the most promising candidates for plasma facing materials in the fusion reactor. Helium (He) in W can influence the retention of hydrogen isotopes. In the present study, W targets were simultaneously irradiated by He+-H 2 + or He+-D 2 + ion beams with the energies of 1 keV or 3 keV, at fixed temperatures in a range of room temperature (R.T.) to 1073 K. Mechanisms of He and hydrogen interaction in W were discussed, especially from the point of He retention, which was characterized by the high-temperature thermal desorption spectroscopy (TDS) and glow-discharge optical emission spectroscopy (GD-OES) measurements. It is found that He desorption shifts to a lower temperature range for the W simultaneously irradiated by 3 keV He+-1 keV H 2 + at 573 K, under He+ fluence up to 1 × 1022 He+m−2. Transmission electron microscope (TEM) observation and annealing treatment at the temperature of 873–1073 K show that the increased He uptake is caused by the formation of dislocation. Enhanced retention amounts for the hydrogen isotopes were also confirmed. Amounts of the dislocation loops introduced by the H 2 +-only irradiation can be reduced by annealing treatment at 873 K, while that introduced by He+ irradiation are quite stable, which grows larger at elevated temperatures. With an increase of H 2 + energy, Helium uptakes at both weak trapping sites and bubbles are increased, while the amounts of hydrogen retention are decreased. It suggests that hydrogen ion has a significant influence on the He trapping sites at the irradiation temperature up to 573 K, while the hydrogen retention is determined by the distribution of He bubbles and dislocation loops. • Helium was mainly trapped by dislocation loops in tungsten irradiated by 3 keV He+- 1 keV H 2 + at 573 K. • Helium concentration in tungsten under 3 keV He+- 1 keV H 2 + irradiation was decreased at 873 K. • Increased hydrogen isotope uptake in tungsten was found when helium was mainly trapped at dislocation loops. [ABSTRACT FROM AUTHOR]

• In order to simulate damage distribution at operation condition, heavy ions were implanted into W. • HIDT simulation code was applied to evaluate the depth profile of D. • The total D retentions were decreased by the damage introduced near the surface region. • The hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface. 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 (D 2 +) 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 D 2 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Biphasic Li 2 TiO 3 –Li 4 SiO 4 tritium breeders with different phase ratios were prepared by the solution combustion synthesis method. The as-prepared biphasic Li 2 TiO 3 –Li 4 SiO 4 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 Li 2 TiO 3 to Li 4 SiO 4. In the case of Li 2 TiO 3 /Li 4 SiO 4 = 2.0, tritium release starts at 400 K, and the tritium migration is mainly controlled by the diffusion process. With the increase of Li 4 SiO 4 content in the biphasic materials, tritium release is influenced by the de-trapping process. • Tritium release dynamics study on neutron irradiated Li 2 TiO 3 –Li 4 SiO 4 breeder. • Tritium release behavior is influenced by the phase ratio. • Tritium recovery at low temperatures seems to be facilitated in biphasic materials. [ABSTRACT FROM AUTHOR]

Abstract 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. [ABSTRACT FROM AUTHOR]