Tritium permeation experiments using reduced activation ferritic/martensitic steel tube and erbium oxide coating

Low concentration tritium permeation experiments have been performed on uncoated F82H and Er₂O₃-coated tubular samples in the framework of the Japan-US TITAN collaborative program. Tritium permeability of the uncoated sample with 1.2 ppm tritium showed one order of magnitude lower than that with 100% deuterium. The permeability of the sample with 40 ppm tritium was more than twice higher than that of 1.2 ppm, indicating a surface contribution at the lower tritium concentration. The Er₂O₃-coated sample showed two orders of magnitude lower permeability than the uncoated sample, and lower permeability than that of the coated plate sample with 100% deuterium. It was also indicated that the memory effect of ion chambers in the primary and secondary circuits was caused by absorption of tritiated water vapor that was generated by isotope exchange reactions between tritium and surface water on the coating.     

Compatibility of erbium oxide coating with liquid lithium–lead alloy and corrosion protection effect of iron layer

Li–Pb compatibility of Er₂O₃ and Er₂O₃-Fe two-layer coatings has been explored for an understanding of corrosion behaviors and effects of the protection layer. The coatings were peeled off after static Li–Pb immersion test at 600 °C due to the degradation of adhesion between the coating–substrate interface. A loss of Er and then subsequent corrosion of Er₂O₃ were shown after immersion at 500 °C for 500 and 1505 h. However, the outer Fe layer played a role to decrease corrosion rate of the coatings by comparing with the results of Er₂O₃ single layer coatings. Deuterium permeation measurements after corrosion tests at 500 °C showed that the Er₂O₃ coatings kept permeation reduction factors of 10²–10³ after 500 h immersion, but seriously degraded after 1505 h immersion. Corrosion mechanisms suggest that corrosion protection properties will be modified by an optimization of the outer Fe layer and a control of oxygen concentration in Li–Pb.     

An advance process of aluminum rich coating as tritium permeation barrier on 321 steel workpiece

We have proposed an advance three-step process, Al-electroplating in ionic liquid followed by heat treating and selectively oxidation, preparing aluminum rich coating as tritium permeation barrier (TPB). In present work, the advance process was applied to 321 steel workpieces. In the Al-electroplating, pieces were coated by galvanostatic electrodeposition at 20 mA/cm² in aluminum chloride (AlCl₃)–1-ethyl-3-methylimidazolium chloride (EMIC) ionic liquid. The Al coating on those pieces all displayed attractive brightness and well adhered to surface of pieces. Within the aluminizing time from 1 to 30 h, a series of experiments were carried out to aluminize 321 steel pieces with Al 20 μm coating at 700 °C. After heat treated for 8 h, a 30 μm thick aluminized coating on piece appeared homogeneous, free of porosity, and mainly consisted of (Fe, Cr, Ni)Al₂, and then was selectively oxidized in argon gas at 700 °C for 50 h to form Al₂O₃ scale. The finally fabricated aluminum rich coating, without any visible defects, had a double-layered structure consisting of an outer γ-Al₂O₃ layer with thickness of 0.2 μm and inner (Fe, Cr, Ni)Al/(Fe, Cr, Ni)₃Al layer of 50 μm thickness. The deuterium permeation reduction factor, PRF, of piece (Φ 80 × 2, L 150 mm) with such coating increased by 2 orders of magnitude at 600–727 °C. The reproducibility of the process was also showed.     

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density) ≥ 0.5 GW/cm² and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1 cm with typical plasma parameters of n_e ≈ 3 × 10²² m^?3 and T_e ≈ 1–2 eV averaged over the plasma lifetime which is below 1 μs. In this study “ITER-Like” mixed deposits with a thickness of about 2 μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2 J/cm² (0.3 GW/cm²) to about 17 J/cm² (2.4 GW/cm²) for 7 ns laser pulses. It is found that the laser energies above about 7 J/cm² (1 GW/cm²) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.     

Preparation of Cr2O3 film by MOCVD as hydrogen permeation barrier

Cr₂O₃ film on structural material as hydrogen permeation barrier can be applied in many areas such as hydrogen storage devices, vacuum solar receivers and fusion reactors. In this study, the Cr₂O₃ film was prepared by MOCVD on 316L stainless steel using chromium(III) acetylacetonate as precursor. The film was characterized by X-ray diffraction (XRD), scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The hydrogen permeation inhibition performance of films was investigated by deuterium permeation experiment. The 366 nm thick Cr₂O₃ film on 316L could reduce the deuterium permeability by 24–117 times at 823–973 K, revealing efficient inhibition to hydrogen permeation. The Cr₂O₃ film is dense, crack-free and has a corundum structure which possesses a more stable structure than a metastable phase or an amorphous phase. Moreover, the crystalline Cr₂O₃ could be easily obtained by MOCVD at a low temperature, e.g. 773 K.     

Investigation of hydrogen concentration and hardness of ion irradiated organically modified silicate thin films

A study of the effects of ion irradiation of organically modified silicate thin films on the loss of hydrogen and increase in hardness is presented. NaOH catalyzed SiNa_wO_xC_yH_z thin films were synthesized by sol–gel processing from tetraethylorthosilicate (TEOS) and methyltriethoxysilane (MTES) precursors and spin-coated onto Si substrates. After drying at 300 °C, the films were irradiated with 125 keV H⁺ or 250 keV N²⁺ at fluences ranging from 1 × 10¹⁴ to 2.5 × 10¹⁶ ions/cm². Elastic Recoil Detection (ERD) was used to investigate resulting hydrogen concentration as a function of ion fluence and irradiating species. Nanoindentation was used to measure the hardness of the irradiated films. FT-IR spectroscopy was also used to examine resulting changes in chemical bonding. The resulting hydrogen loss and increase in hardness are compared to similarly processed acid catalyzed silicate thin films.     

Heteroepitaxial Er0.49Gd0.51Si1.7 layers formed by channeled ion beam synthesis

Epitaxial ternary silicide Er_0.49Gd_0.51Si_1.7 layers with a good crystalline quality (χ_min of Er and Gd is 3.7%) have been formed by 60 keV Er and Gd ion implantation into Si(1 1 1) substrates to a total dose of 1.0 × 10¹⁷/cm² at 450°C using channeled ion beam synthesis (CIBS). The composition, the structure, the strain and the thermal stability of these layers have been studied using energy dispersive spectroscopy (EDS), Rutherford backscattering (RBS)/channeling and X-ray diffraction (XRD). It is shown that the perpendicular and parallel elastic strains of the Er_0.49Gd_0.51Si_1.7 epilayer are e^⊥=−0.46% ± 0.02% and e⁶=+0.73% ± 0.19%. The layer is stable up to 900°C. Annealing at 950°C results in a phase transformation.     

Repulsive potentials for low energy projectiles incident on copper surfaces

Interatomic potentials that are relevant for low energy (<2 keV) projectiles (X = He, Cl, Ar, Cu) incident on the Cu(1 0 0) and Cu(1 1 1) surfaces (represented by Cu₁₀ clusters), and on isolated Cu atoms, have been calculated from first principles using density functional theory. For energies above 10–20 eV, the diatomic X–Cu potentials provide acceptable approximations (typically within 1–2 eV) to the corresponding potentials for the surface environment.     

Oxidation of Zircaloy-4 by oxygen and the production of water

The interaction of isotopic oxygen (¹⁸O₂) with Zircaloy-4 (Zry-4) at 150 and 300 K has been studied using Auger electron spectroscopy (AES) and temperature-programmed desorption (TPD) methods. AES reveals the oxidation of the Zry-4 surface, reflected in shifts of the Zr(MNV) and Zr(MNN) features by about 5.5 and 3.0 eV, respectively, for both adsorption temperatures. The O(KLL)/Zr(MNN) Auger peak-to-peak height ratios as a function of exposure show the same trends at both temperatures. Following ¹⁸O₂ adsorption at 150 or 300 K, TPD experiments show hydrogen desorption near 400 K that is attributed to the presence of a surface-stabilized form of hydrogen. Additionally, water (H₂¹⁸O and H₂¹⁶O) desorption below 200 K and above 700 K is observed after 150 K oxygen adsorption. However, after oxygen adsorption at 300 K the only significant desorption features are from isotopic water (H₂¹⁸O). These findings indicate that mass transport involving the near-surface region contributes to the observed desorption, and that this behavior is dependent on the original adsorption temperature. Charging experiments using D₂ prior to and after ¹⁸O₂ adsorption were also performed and support our conclusions about the role of surface–subsurface mass transport in this system.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

On the study of catalytic membrane reactor for water detritiation: Membrane characterization

Tritium waste recycling is a real economic and ecological issue. Generally under the non-valuable Q₂O form (Q = H, D or T), waste can be converted into fuel Q₂ for a fusion machine (e.g. JET, ITER) by isotope exchange reaction Q₂O + H₂ = H₂O + Q₂. Such a reaction is carried out over Ni-based catalyst bed packed in a thin wall hydrogen permselective membrane tube. This catalytic membrane reactor can achieve higher conversion ratios than conventional fixed bed reactors by selective removal of reaction product Q₂ by the membrane according to Le Chatelier's Law.This paper presents some preliminary permeation tests performed on a catalytic membrane reactor. Permeabilities of pure hydrogen and deuterium as well as those of binary mixtures of hydrogen, deuterium and nitrogen have been estimated by measuring permeation fluxes at temperatures ranging from 573 to 673 K, and pressure differences up to 1.5 bar. Pure component global fluxes were linked to permeation coefficient by means of Sieverts’ law. The thin membrane (150 μm), made of Pd–Ag alloy (23 wt.%_Ag), showed good permeability and infinite selectivity toward protium and deuterium. Lower permeability values were obtained with mixtures containing non permeable gases highlighting the existence of gas phase resistance. The sensitivity of this concentration polarization phenomenon to the composition and the flow rate of the inlet was evaluated and fitted by a two-dimensional model.     

Permeation measurements for investigating atomic hydrogen flux and wall pumping/fuelling dynamics in QUEST

In order to investigate the overall atomic hydrogen background and the dynamic characteristics of wall pumping/fuelling phenomenon, a permeation probe system has been developed and applied in the spherical tokamak QUEST. Reliability of measurements, within ±3% accuracy and a positive correlation with the hydrogen line emission over three orders of magnitude have been demonstrated for more than 3000 various plasma discharges. By comparison of the experimental permeation (flux) curves with the numerically simulated curves, the net incident atomic hydrogen flux is evaluated in the range of 1 × 10¹⁹ H m^?2 s^?1 to 4 × 10²⁰ H m^?2 s^?1. The atomic flux has been investigated as a function of various plasma operation parameters like RF power, gas pressure and magnetic configuration. Using the static particle balance and permeation measurements, the progress in wall conditioning has been investigated. An inverse correlation between the atomic hydrogen flux and improvement in wall pumping has been observed over the two campaigns.     

Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose.Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min^?1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 10¹⁶ Bq kg^?1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water.The combination of the μCCR with the O₂ sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R&D activity on HTW processing with PERMCAT.     

A calorimetric study of high temperature phase stability in Fe–U alloys

A differential scanning calorimetry study of high temperature phase equilibria and phase transformation kinetics in Fe_100?xU_x binary alloys, with x varying from 0 to 95 mass% U is undertaken. Accurate measurement of transformation temperatures pertaining to: (i) α-Fe → γ-Fe → δ-Fe polymorphic phase change, (ii) UFe₂ + γ-Fe → L and U₆Fe + UFe₂ → L transformations and (iii) melting has been made as a function of uranium concentration. The measured transformation temperatures are used to construct the binary Fe?U phase diagram, which showed general agreement with the latest assessment. The L → UFe₂ + γ-Fe eutectic reaction is found to occur at 1357 ± 5 K, with the eutectic composition of 47 mass%. The heat of transformation associated with this invariant reaction is estimated to be 19,969 ± 1736 J mol of atoms^?1. The L → U₆Fe + UFe₂ reaction occurs at 89 mass%, and at 1001 ± 5 K, with a heat of transformation 20,250 ± 2113 J mol of atoms^?1. The heat of melting of stoichiometric UFe₂ is estimated to be 20,983 ± 2098 J mol of atoms^?1, which is higher than the currently assessed value by 30%. A non-faceted microstructural morphology is found to accompany the eutectic solidification process of all the alloy compositions.     

Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1], [2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.     

Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 °C) and the beam-off (20 °C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 × 10⁴ MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environment.     

Erbium environment on Er-doped silica and alumino-silicate glass films: An EXAFS study

Er-doped dielectric films are characterized by the emission of a photoluminescence signal at λ = 1.54 μm, the main used in the optical telecommunications. The efficiency of the radiative emission is strongly related to the characteristics of the Er³⁺ environment. Er-doped SiO₂ films (synthesized by rf-magnetron co-sputtering) and 87SiO₂:10Al₂O₃:3Na₂O silicate glass films doped with 0.5 mol% of Er (prepared by sol–gel route and subsequently doped with silver by Ag⁺ ? Na⁺ field-assisted solid-state ion exchange) were studied by extended X-ray absorption fine structure spectroscopy performed at Er L_III-edge (Italian beamline GILDA of the ESRF). In the silica samples the Er coordinates about 4.5 O atoms at a short distance (R = 2.07–2.13 Å), similar to the one observed in Er-doped glasses when the preparation conditions are far from the thermodynamical equilibrium. In alumino-silicate samples the first shell of atoms is formed of 5.5–7.5 O atoms at a distance of about 2.31 Å, showing a local structure similar to other Er-doped sol–gel glasses and glass–ceramics. A comparison between the first shell structure around Er ions and the different intensity of the photoluminescence emission suggests that the increase of the radiative emission upon thermal annealing is mainly related to the decrease of the defects number in the glass structure as a consequence of the annealing.     

Thermophysical properties of U2Mo intermetallic

The current paper has investigated specific heat capacity, coefficient of thermal expansion and phase transition temperatures of U₂Mo intermetallic. The transformation of U₂Mo phase to bcc (γ-uranium) phase occurred at 853 K. The coefficient of thermal expansion has been determined in the temperature range 423–873 K to be 14.65 × 10^?6 K^?1. The specific heat data showed a smooth curve up to 773 K and above it a change in trend to follow a λ-shape was observed. The specific heat values of U₂Mo have been found to be lower than that calculated from the additivity rule. This study provides first time data on U₂Mo. The data obtained from this investigation were compared to available literature on U–Mo alloys.     

Study and optimization of boronization in Alcator C-Mod using the Surface Science Station (S3)

A Surface Science Station (S³) on the Alcator C-Mod tokamak is used to study and optimize the location and rate of boron film deposition in situ during electron cyclotron (EC) discharge plasmas using 2.45 GHz radio-frequency (RF) heating and a mixture of helium and diborane (B₂D₆) gasses. The radial profile of boron deposition is measured with a pair of quartz microbalances (QMB) on S³, the faces of which can be rotated 360° including orientations parallel and perpendicular to the toroidal magnetic field B_T ～0.1 T. The plasma electron density is measured with a Langmuir probe, also on S³ in the vicinity of the QMBs, and typical values are ～1 × 10¹⁶ m^?3. A maximum boron deposition rate of 0.82 μg/cm²/min is obtained, which corresponds to 3.5 nm/min if the film density is that of solid boron. These deposition rates are sufficient for boron film applications between tokamak discharges. However the deposition does not peak at the EC resonance as previously assumed. Rather, deposition peaks near the upper hybrid (UH) resonance, ～5 cm outboard of the EC resonance. This has implications for RF absorption, with the RF waves being no longer damped on the electrons at the EC resonance. The previously inferred radial locations of critical erosion zones in Alcator C-Mod also need to be re-evaluated. The boron deposition profile versus major radius follows the ion flux/density profile, implying that the boron deposition is primarily ionic. The application of a vertical magnetic field (B_V ～0.01 T) was found to narrow the plasma density and boron deposition profiles near the UH resonance, thus better localizing the deposition. A Monte Carlo simulation is developed to model the boron deposition on the different QMB/tokamak surfaces. The model requires a relatively high boron ion gyroradius of ～5 mm, indicating a B⁺¹ ion temperature of ～2 eV, to match the deposition on QMB surfaces with different orientation to B_T. Additionally, the boron ion trajectories become de-magnetized at high neutral gas throughput (～0.5 Pa m³ s^?1) and pressure (～2 Pa) when the largest absolute deposition rates are measured, resulting in deposition patterns, which are independent of surface orientation to B_T in optimized conditions.     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.