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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...

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.

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.

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.

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.

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.

Double strand breaks (DSBs) of DNA molecules in tritiated water was examined under sterilized and non-sterilized conditions using a single molecule observation method. The genome DNA of bacteriophage T4 GT7 was immersed in sterilized tritiated water (5.2 MBq/cm3) and non-sterilized tritiated water (4.2 MBq/cm3) for 1, 7 and 14 day(s). Then the length of DNA molecules was measured using a fluorescence microscope after intercalation of fluorescent dye. The dose rate was 1.4–1.7 × 10−2 Gy/h and the dose level was 0.41–5.8 Gy. The rate of DSBs induced by β-rays from tritium was successfully evaluated under the sterilized conditions and the value comparable with the DSB rate under γ-ray irradiation (Noda et al., Scientific Reports 7 (2017) 8557) was obtained. The length of DNA molecules in non-sterilized tritiated water was clearly shorter than that in the sterilized tritiated water. This observation suggested that the effects of tritium was far weaker than that of microorganisms (e.g. bacteria) and impurities in water even at the tritium concentration as high as 5.2 MBq/cm3.

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.

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.

The number of DNA double strand breaks (DSBs) of genome-sized DNA was quantitatively evaluated under several types of radiation sources. The characteristics of radiation damages by different linear energy transfer (LET) and the effect of radiation protection by tea catechins were also studied as a function of its concentration. After β-rays and γ-rays irradiation to samples, the length of genome size DNA molecules (bacteriophage T4 GT7 DNA; 166 kbp) was measured by single molecule observation method using fluorescence microscope, which can estimate DSBs quantitatively. It was found that the number of DSBs was increased with increasing LET due to high radical density. By addition of EGCg ((-)-epigallocatechin gallate), the number of DSBs was reduced with a small concentration of 1 µM.

Dependence of irradiation temperature on tritium retention has been studied using stainless steel type 316L (SS316L) as a model sample. A mixture of D2+ and DT+ ions was used and two kinds of ion energy, 0.5 and 2.5 keV, were applied. In case of irradiation tests by 0.5 keV, tritium retention decreased with increasing temperature up to 523 K, while above this temperature it contrarily showed an increase tendency. Such a concave change was not observed for irradiation tests at 2.5 keV. The retention was almost same until 400 K, but above this temperature it decreased gradually. It was seen from the analyses by X-ray photoelectron spectroscopy that most of surface was initially covered with the carbon and oxygen species at room temperature. Among metallic elements, constituents such as Fe and Ni were metallic states more than 60 % at room temperature, while metallic chromium atoms were little observed. Both fractions of the metallic chromium and iron atoms in the major base metals of SS316 L increased with an increase in temperature, but metallic nickel atoms relatively decreased. It was suggested, therefore, that real surface states of the irradiation materials play an important role for behavior of tritium retention.

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.

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.

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...

The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the respo...

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...

Understanding and managing D retention in plasma facing components is essential for tritium safety in fusion reactors. Neutron irradiated and virgin low carbon arc cast (LCAC) Mo, as well as Mo foil samples with and without He pre-irradiation, were used to investigate D retention. D and He retention were investigated simultaneously in thermal desorption spectroscopy using a high resolution residual gas analyzer. Results show a significant increase in D retention with He pre-irradiation. Vacancies and vacancy clusters are found to retain D in LCAC samples, but neutron irradiated Mo retains more D in vacancy clusters.

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.

Molten salt electrodeposition is a promising technology to manufacture the first wall of a fusion reactor. Deuterium (D) retention behavior in molten salt electrodeposition tungsten (W) coatings has been investigated by D-plasma exposure in the EAST tokamak and D-ion implantation in an ion beam facility. Tokamak exposure experiments demonstrate that coatings prepared with lower current density exhibit less D retention and milder surface damage. Deuterium-ion implantation experiments indicate the D retention in the molten salt electrodeposition W is less than that in vacuum plasma spraying W and polycrystalline W.

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.

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

Deuterium retention behavior in tungsten damaged by fast neutrons at high temperatures (0.43 dpa at 918 K and 0.74 dpa at 1079 K) and 6.4 MeV Fe2+ (0.3 dpa at R.T.) were investigated to evaluate the tritium retention property of fusion reactor divertors. A deuterium gas absorption method was carried out to avoid additional damage that may be induced by plasma exposure, then, deuterium retention and desorption behaviors were investigated quantitatively by means of thermal desorption spectroscopy and the following simulation code. The deuterium desorption spectra for tungsten samples were analyzed by the numerical code which includes the elementary steps of hydrogen isotope migration processes including diffusion, trapping, detrapping, and surface recombination. The evaluated deuterium detrapping energy from the irradiation defects in neutron irradiated tungsten sample was larger than that in 6.4 MeV Fe2+ irradiated tungsten. It was suggested that the dominant deuterium trapping site in the neutron irradiated tungsten would be voids which was formed by the accumulation of vacancies during neutron irradiation under high temperature and long duration.

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.

To address the transport and trapping of hydrogen isotopes, several permeation experiments are being pursued at both Sandia National Laboratories (deuterium gas-driven permeation) and Idaho National Laboratories (tritium gas- and plasma-driven tritium permeation). These experiments are in part a collaboration between the US and Japan to study the performance of tungsten at divertor relevant temperatures (PHENIX). Here we report on the development of a high temperature (≤1150 °C) gas-driven permeation cell and initial measurements of deuterium permeation in several types of tungsten: high purity tungsten foil, ITER-grade tungsten (grains oriented through the membrane), and dispersoid-strengthened ultra-fine grain (UFG) tungsten being developed in the US. Experiments were performed at 500–1000 °C and 0.1–1.0 atm D2 pressure. Permeation through ITER-grade tungsten was similar to earlier W experiments by Frauenfelder (1968–69) and Zaharakov (1973). Data from the UFG alloy indicates marginally higher permeability (

Deuterium (D) plasma driven permeation (PDP) experiments for tungsten (W) samples were conducted by a linear radio frequency (RF) plasma device. In the PDP experiment, the W sample surface is perpendicular to the grain elongation direction. The D ion flux is in the order of 1021 m−2 s−1 determined by a double Langmuir probe. The results show that bias had a limited effect on the D plasma driven permeation behavior. The D recombination coefficient on W surface is obtained at the temperature ranging from 740 K to 1031 K. The experimentally measured recombination coefficient for a pristine W surface is lower than that for a clean W surface. The effect of recombination coefficient on the D permeation and retention behaviors in W are studied by Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code. The low recombination coefficient leads to a high D concentration in W and a high permeation rate at the back surface. The D permeation rate through a 8 mm thick W with a pristine surface is estimated to be 8.1 × 1018 D m s−1 under the incident ion flux of 1 × 1024 m−2 s−1 and temperature of 1173 K.

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.

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...

The deuterium retention behavior for the Vacuum Plasma Spraying (VPS) tungsten (W) coating was studied to demonstrate the tritium retention as a function of heating temperature. It was found that t...

The effect of carbon implantation for the dynamic recycling of deuterium, which demonstrates tritium recycling, including retention and sputtering, was investigated using in-situ sputtered particle...

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.

To investigate the behavior of hydrogen on a water–metal boundary, a series of experiments have been performed that studied tritium permeation into a pressurized water jacket through pure iron piping, which contained approximately 1 kPa of pure tritium gas at 423 K, while monitoring the chemical forms of tritium. Additionally, the behavior of deuterium, which was generated on the heavy water–metal boundary and transferred to the metal, was also investigated using a pressure-proof heavy water vessel. Actual deuterium transfer through various metal piping, such as Fe, Ni, SS304, etc., was detected clearly by QMS at 423–573 K. Moreover, using the above heavy water system, we have succeeded in detecting hydrogen isotopes simultaneously transferring from and to the metal surface by introducing hydrogen gas to the Ni piping after deuterium permeation from the heavy water side had stabilized.

To evaluate hydrogen isotope retention enhancement in W by plasma exposure, the stress relieved tungsten samples were placed at three or four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (Higher heat load area) and ER (erosion dominated area) during 2011 (15th) or 2012 (16th) plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼6700 shots of hydrogen plasma in a 2011 plasma experiment campaign and ∼5000 shots in a 2012 plasma campaign. Thereafter, additional 1.0 keV deuterium ion implantation was performed to evaluate the change of hydrogen isotope retention capacity by plasma exposure. It was found that more than 50 times of hydrogen retention enhancement for DP sample was derived compared to that for pure W. In especially, the carbon-dominated mixed-material layer would control the hydrogen isotope retention for all the area except for erosion-dominated area, indicating that the chemical structure for carbon-dominant mixed-material layer would govern the H and D retention enhancement for most area by long-term plasma exposure. Therefore, the surface area for carbon material would be one of key issues for the determination of hydrogen isotope retention in first wall, even if all tungsten first walls will be used.

The deuterium trapping behaviors in tungsten damaged by light ions with lower energy (10 keV C+ and 3 keV He+) or a heavy ion with higher energy (2.8 MeV Fe2+) were compared by means of TDS to understand the effects of cascade collisions on deuterium retention in tungsten. By light ion irradiation, most of deuterium was trapped by vacancies, whose retention was almost saturated at the damage level of 0.2 dpa. For the heavy ion irradiation, the deuterium trapping by voids was found, indicating that cascade collisions by the heavy ion irradiation would create the voids in tungsten. Most of deuterium trapped by the voids was desorbed in higher temperature region compared to that trapped by vacancies. It was also found that deuterium could accumulate in the voids, resulting in the formation of blisters in tungsten.

The release behaviors of hydrogen isotopes in Pb-16Li introduced by thermal gas exposure or produced by thermal neutron irradiation were compared to investigate hot atom reactions of tritium. The solubility of hydrogen isotope was evaluated to be S = 6.56 × 10 −7 exp(−0.11 [eV]/kT) [at. fr, Pa 0.5 ], which is consistent with the literature values. The tritium TDS spectra for Pb-16Li with thermal neutron irradiation showed that about 5% of tritium was released at single release stage around 600 K, which was higher than the melting point of Pb-16Li, although no release peak was found at around 600 K for the hydrogen isotope-doped Pb-16Li. The kinetic analysis indicated that the tritium release would be associated with the dissociation of Li-T bond, indicating that LiT was formed in Pb-16Li by thermal neutron irradiation, suggesting that the formation of Li-T bonds should be considered to estimate tritium retention in Pb-16Li eutectic blanket systems in fusion reactors.

The dependence of irradiation defect density on hydrogen isotope release behaviors in Li2TiO3 was studied. Li2TiO3 was exposed to gaseous deuterium and, thereafter, gamma-ray irradiation was performed with various gamma-ray doses to change the density of irradiation defects. The deuterium release behavior was measured by TDS. The density of the defects and the state of O D bonds in the sample were elucidated by ESR and FT-IR, respectively. Most of deuterium was adsorbed on the surface or trapped by intrinsic defects after deuterium gas exposure. However, O D bonds were increased as the gamma-ray dose was increased. In addition, the irradiation defects like E-center, O−-center and O2−-center were observed in gamma-ray irradiated samples. This indicated that the density of irradiation defect control the deuterium stable trapping by oxygen. These facts conclude that tritium release temperature will be shifted toward higher temperature as the operation time increased and irradiation defects are accumulated.

The D 2 + fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660 K in the TDS spectrum for a sample implanted with D 2 + up to the fluence of 10 23 D + m −2 , which desorption stage was not observed the D 2 + implanted sample with the fluence less than 10 22 D + m −2 . The TEM observation showed that the highly dense voids were formed in tungsten by D 2 + implantation with the fluence of 10 23 D + m −2 , considering that the D would be trapped by voids. To understand the D trapping by voids in C + implanted tungsten, C + –D 2 + sequential implantation experiments at various C + implantation temperatures were performed. It was found that the amount of D desorbed around 560 K was increased by increasing the C + implantation temperature. The formation of the voids was observed with increasing the C + implantation temperature by TEM, indicating that the increase of D desorption around 560 K was caused by the formation of voids. However, the desorption temperature of D trapped by voids in C + implanted sample was lower than that in D 2 + implanted one. TEM observation and XPS measurement indicated that this difference was caused by the increase of void size and/or the presence of implanted carbon.

The stress relieved tungsten samples were placed at three positions, PI (sputtering erosion dominated area), DP (deposition dominated area) and HL (Higher heat load area) during 15th plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼ 6700 shots of hydrogen plasma in a 15th long-term experiment campaign in LHD. Thereafter, the additional deuterium ion implantation to these tungsten samples was performed to evaluate the change of hydrogen isotope retention capacity in the samples by long-term plasma exposure. It was found that the carbon-dominant mixed-material layer with more than 100 nm thickness was formed on a wide area of the tungsten surface. The thicker mixed-material layer was formed on the DP sample, where the deuterium retention was about 21 times as high as that for pure W. The major desorption temperature of deuterium was shifted toward higher temperature side, which was comparable to the trapping characteristic of carbon or irradiation damages.

1National Institute for Fusion Science, Toki 509-5292, Japan 2Sandia National Laboratories, Albuquerque, NM 87123-1129, USA 3Shimane University, Matsue 690-8504, Japan 4University of Californian, San Diego, CA 92093-0417, USA 5Kyushu University, Fukuoka 812-8581, Japan 6Shizuoka University, Shizuoka 422-8529, Japan 7Department of Materials Science and Engineering, University of Tennessee, Knoxville, USA 8Institute of Advanced Energy, Kyoto University, Kyoto 611-0011, Japan 9Pacific Northwest National Laboratory, Richland, WA 99352, USA 10Institute of Laser Engineering, Osaka University, Suita 5650871, Japan

To investigate the effects of neutron irradiation on hydrogen isotope retention in tungsten, disk-type specimens of pure tungsten were irradiated in the High Flux Isotope Reactor in Oak Ridge National Laboratory followed by exposure to high flux deuterium (D) plasma in Idaho National Laboratory. The results obtained for low dose n-irradiated specimens (0.025dpa for tungsten) are reviewed in this paper. Irradiation at coolant temperature of the reactor (around 50°C) resulted in the formation of strong trapping sites for D atoms. The concentrations of D in nirradiated specimens were ranging from 0.1 to 0.4mol% after exposure to D plasma at 200 and 500°C and significantly higher than those in nonirradiated specimens because of D-trapping by radiation defects. Deep penetration of D up to a depth of 50­100µm was observed at 500°C. Release of D in subsequent thermal desorption measurements continued up to 900°C. These results were compared with the behaviour of D in ion-irradiated tungsten, and distinctive features of n-irradiation were discussed. [doi:10.2320/matertrans.MG201204]