Tritium Imaging Plate Research Articles

A comparative study on the hydrogen dissolution and release behaviors in the zirconate proton conductors by TDS and TMAP4 analysis

Proton-conducting oxides are potential materials for electrochemical devices such as fuel cells, hydrogen pumps, hydrogen sensors, and the tritium purification and recovery system in nuclear fusion reactors. The hydrogen concentration in oxide materials is important, but its precise measurement is difficult. In this study, thermal desorption spectroscopy (TDS) was used to investigate hydrogen dissolution and release behavior in proton-conducting oxides, yttrium (Y), and cobalt (Co) doped barium-zirconates in the temperature range of 673-1273 K using deuterium (D2) and heavy water (D2O). Specimens were prepared with conventional powder metallurgy: the powder of three zirconates, BaZr0.9Y0.1O3-a (BZY), BaZr0.955Y0.03Co0.015O3-a (BZYC), and CaZr0.9In0.1O2.95 (CZI) was pressed into discs and fired in the air at 1873 K for 20 h. The densities of the sintered BZY, BZYC, and CZI specimens were 98 percent, 99.7 percent, and 99.5 percent of the theoretical densities respectively. XRD, SEM, and EDX were performed for the structural and morphological analysis of the sintered samples. From TDS measurement, a similar trend of temperature-dependent hydrogen solubility was obtained for all samples compared to the tritium imaging plate (TIP) methods literature data of HT- and DTO-exposed samples. To compare the experimental results of the deuterium desorption profile derived by TDS analysis, the simulation code of the tritium migration analysis program, version 4 (TMAP4) was employed.

An overview of tritium retention in dust particles from the JET-ILW divertor

Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles.

Comparative Studies of Hydrogen Dissolution and Release Behavior in Zirconate Oxides by TDS and TMAP4 Analysis

Proton-conducting oxides are potential materials for electrochemical devices such as fuel cells, hydrogen pumps, hydrogen sensors, and the tritium purification and recovery system in nuclear fusion reactors. The hydrogen concentration in oxide materials is important, but its precise measurement is difficult. In this study, thermal desorption spectroscopy (TDS) was used to investigate hydrogen dissolution and release behavior in proton-conducting oxides, yttrium (Y), and cobalt (Co) doped barium-zirconates in the temperature range of 673-1273 K using deuterium (D2) and heavy water (D2O). Specimens were prepared with conventional powder metallurgy: the powder of three zirconates, BaZr0.9Y0.1O3-a (BZY), BaZr0.955Y0.03Co0.015O3-a (BZYC), and CaZr0.9In0.1O2.95 (CZI) was pressed into discs and fired in the air at 1873 K for 20 h. The densities of the sintered BZY, BZYC, and CZI specimens were 98 percent, 99.7 percent, and 99.5 percent of the theoretical densities respectively. XRD, SEM, and EDX were performed for the structural and morphological analysis of the sintered samples. From TDS measurement, a similar trend of temperature-dependent hydrogen solubility was obtained for all samples compared to the tritium imaging plate (TIP) methods literature data of HT- and DTO-exposed samples. To compare the experimental results of the deuterium desorption profile derived by TDS analysis, the simulation code of the tritium migration analysis program, version 4 (TMAP4) was employed.

Evaluation of the hydrogen solubility and diffusivity in proton-conducting oxides by converting the PSL values of a tritium imaging plate

Proton-conducting oxides have potential applications in hydrogen sensors, hydrogen pumps, and other electrochemical devices including the tritium purification and recovery systems of nuclear fusion reactors. Although the distribution of hydrogen (H) in such oxide materials is an important aspect, its precise measurement is difficult. In the present study, the hydrogen solubility and diffusivity behavior of BaZr0.9Y0.1O2.95 (BZY), BaZr0.955Y0.03Co0.015O2.97 (BZYC), and CaZr0.9In0.1O2.95 (CZI) were studied using tritiated heavy water vapor i.e., DTO (~2 kPa, tritium (T) = 0.1%) by converting the photostimulated luminescence (PSL) values of the imaging plate (IP). The samples were exposed to DTO vapor at 673 K for 2 h or at 873 K for 1 h. The disc-shaped oxide specimens (diameter ~7.5 mm; thickness ~2.3 mm; theoretical density (TD) > 98 %) were prepared by conventional powder metallurgy. The IP images of the specimen surfaces of all the three materials T-exposed revealed that BZY showed the most uniform T distribution with the highest tritium activity. The cross-sectional T concentration profiles of the cut specimens showed that T diffused deeper into BZY and BZYC than into CZI. The hydrogen solubility and diffusivity in the CZI specimen were lower than that in the BZY and BZYC specimens. This suggested that barium zirconates were more favorable proton conductors than calcium zirconates.

Tritium removal from the LHD first-wall by the hydrogen plasma discharge

Abstract During the 19th experimental campaign in the Large Helical Device (LHD) in 2017, deuterium (D) plasma experiments were performed for the first time in the LHD. Tritons were generated by the D-D reaction in the core plasma during the plasma discharges. To remove the tritium in plasma facing surfaces for safety in maintenance, hydrogen (H) plasma experiments were performed for one month after the four months of D plasma experiments. In this study, the removal effect of the tritium by H plasma discharges was investigated. Six stainless steel (SUS316L) samples were introduced into a position of the first-wall of the LHD in the last month of the D experiments, which were exposed to the D plasma for one month and then, three samples were continued to exposure to the following H plasma for another one month. The experimental results showed that one-half of the tritium in the surface region of the SUS sample was removed, which was measured by the TIPT (Tritium imaging plate technique). Further, hydrothermal treatments were performed and the results showed that the tritium retention in the bulk of the SUS materials was also reduced. Meanwhile, the carbon content in the near-surface region was also reduced after H plasma experiments, which reduces the possibility of the formation of the C H chemical bond.

Hydrogen Isotope Dissolution and Release Behavior of Rare Earth Oxides

Hydrogen release behavior from rare earth oxides (REOs) (Y2O3, Sm2O3, Eu2O3, Gd2O3, Dy2O3, Er2O3, and Yb2O3) exposed to 133 Pa of deuterium (D2) gas or 2 kPa of heavy water (D2O) vapor at 873 K for 5 h was examined using thermal desorption spectroscopy. Hydrogen solubility and diffusivity in Y2O3, Gd2O3, Dy2O3, Er2O3, and Yb2O3 exposed to a deuterium-tritium gas mixture (5% to 7% T, 133 Pa) at 873 K and 973 K for 5 h were determined using a tritium imaging plate method. The structural and morphological properties of sintered disk specimens of those REOs were evaluated using an X-ray diffractometer and a scanning electron microscope. From the obtained results, the REO materials were clearly categorized into two kinds in terms of their crystal structure and hydrogen solubility: Monoclinic specimens of Sm2O3, Eu2O3, and Gd2O3 had relatively high hydrogen solubility and diffusivity, while cubic Y2O3, Dy2O3, Er2O3, and Yb2O3 had lower ones. The present study suggests that the cubic REOs could be suitable in a nuclear fusion reactor as the tritium barrier materials.

Investigation of remaining tritium in the LHD vacuum vessel after the first deuterium experimental campaign

Remaining tritium in the vacuum vessel after the first deuterium plasma experimental campaign conducted over four months was investigated in the large helical device (LHD) for the first time in stellarator/heliotron devices by using the tritium imaging plate technique. In-vessel components such as divertor tiles and first wall panels, and long-term material probes retrieved from the vacuum vessel were analyzed. The in-vessel component in which tritium remained most densely is the baffle part of divertor tiles made of graphite retrieved from the inboard-side divertor. Asymmetric tritium retention is observed on divertor tiles located at magnetically symmetric positions, and can be attributed to the toroidal field direction dependence of the asymmetric loss of energetic tritons generated by deuterium–deuterium nuclear fusion reactions. On the first wall, tritium remained in a deposited layer, which mainly consists of carbon.

Effects of activation products on tritium measurements for tungsten in fusion reactors

Influence of neutron activated products (NAPs) is one of the primary problems for tritium measurements in tungsten as plasma-facing materials (PFMs). Both Monte Carlo simulation was introduced to calculate the amount of NAPs after different decay time and at different location of the divertor in detail. Then, the effects of NAPs on the commonly used tritium measurement methods for PFMs, including chemical etching method, β-ray induced X-ray spectrometry, tritium imaging plate technique and tritium calorimeter, have been evaluated quantitatively for PFM samples with different tritium depth profile, sample thickness and sample cooling time. Results indicate that more than one hundred kinds of NAPs would generate through the irradiation of fusion neutrons and NAPs with half-lives shorter than 1E-03 year and large activities, such as W-187, W-183 m etc., account for a substantial part of total activity directly after exposure and their amounts rapidly decrease to negligible as decay time increases. However, within cooling time from 3 to 60 months, the influence of tritium measurements is barely dominated by W-185, W-181, Ta-182 and Ta-179. In addition, it is found that the influence of NAPs on chemical etching method is the minimum while the other three approaches are still strongly affected by NAPs after several months. And it usually takes several years to let the signal from NAPs become acceptable comparing with signal from tritium in measurements.

Tritium retention characteristics in dust particles in JET with ITER-like wall

A tritium imaging plate technique (TIPT) in combination with an electron-probe microscopic analysis (EPMA) were applied to examine tritium (T) retention characteristics in individual dust particles collected in the Joint European Torus with the ITER-like Wall (JET-ILW) after the first campaign in 2011–2012.A lot of carbon (C)-dominated dust particles were found, which would be pre-existing carbon deposits in the JET-C or released carbon particles from the remaining carbon-fiber components in the JET-ILW. Most of T was retained at the surface of and/or in the C-dominated dust particles. The retention in tungsten, beryllium and other metal-dominated dust particles is relatively lower by a factor of 10–100 in comparison with that in the C-dominated particles.

Using a Tritium Imaging Plate Technique to Measure the Hydrogen Solubility and Diffusivity of Zr-Doped BaInO2.5

A tritium imaging plate technique (TIPT) was used to measure the hydrogen solubility and diffusivity of a proton-conducting material, Zr-doped BaInO2.5. After being exposured to tritiated water vapor, the specimen was cut into halves and photo-stimulated luminescence (PSL) images of the cross-section of the cut specimen were obtained. Based on the PSL distribution profiles and intensities, hydrogen diffusivity and solubility in the specimens were determined. The hydrogen solubility and diffusivity in Zr-doped BaInO2.5 clearly depended on the content of Zr. This characteristic of the Zr content is in agreement with the Zr-content dependency of its proton conductivity, which was measured by an alternating current (AC) impedance method.

Release behavior of tritium in pure copper and its alloys into pure water at ambient temperature

Release behaviors of tritium (T) from pure copper (Cu) and its alloys into pure water were examined at ambient temperature by a tritium imaging plate technique and a liquid scintillation counting technique. Two mechanisms govern the liberation of T into pure water; one is rapid release and the other is chronic release. The former is caused by diffusional release of T dissolved in bulk of Cu alloys and the latter by release of T strongly bound on/in surface oxide layers.

Tritium retention in individual metallic dust particles examined by a tritium imaging plate technique

Tritium imaging plate technique (TIPT) has been applied to examine tritium (T) retention in individual particles made of titanium (Ti) with 30 and 100 μm in diameter and tungsten (W) with 50 μm in diameter. Distribution of T radioactivity observed by TIPT corresponded well to spatial distribution of the particles. In a limited case of uniform and high T concentration in the bulk of the individual particle, the amount of T is directly quantified from T radioactivity by a master curve method. Density and size of the particle and T concentration profiles in the bulk of the particle are important factors to change emission behavior of T β-ray and thus accurate quantification of the amount of T in the individual particle.

Hydrogen solubility and diffusivity in a barium cerate protonic conductor using tritium imaging plate technique

A tritium imaging plate technique has been applied to visualize hydrogen distribution and examine hydrogen solubility and diffusivity in a proton-conducting oxide, Y-doped BaCeO3 (BaCe0.9Y0.1O3−α). Tritium charging of the BaCe0.9Y0.1O3−α specimens was carried out by a gas absorption method using partially tritiated water vapor (HTO, 3 kPa, T/H~10−6) at temperatures ranging from 673K to 873K for a given time. After charging, tritium distributions of the surface and cross section of the halved specimens were visualized using an imaging plate technique. From the tritium concentration and distributions of the surface and cross section, hydrogen solubility and hydrogen (tritium) diffusivity of the BaCe0.9Y0.1O3−α specimens were determined.

Determination of Hydrogen Diffusion Coefficients in F82H by Hydrogen Depth Profiling with a Tritium Imaging Plate Technique

Hydrogen diffusion coefficients in a reduced activation ferritic/martensitic steel (F82H) and an oxide dispersion strengthened F82H (ODS-F82H) have been determined from depth profiles of plasma-loaded hydrogen with a tritium imaging plate technique (TIPT) in the temperature range from 298 K to 523 K.Data of hydrogen diffusion coefficients, D, in F82H are summarized as D [m2 s−1] =1.1×10−7 exp(-16[kJ mol−1]/RT). The present data indicate almost no trapping effect on hydrogen diffusion due to an excess entry of energetic hydrogen by the plasma loading, which results in saturation of the trapping sites at the surface and even in the bulk. In the case of ODS-F82H, data of hydrogen diffusion coefficients are summarized as D [m2 s-1] =2.2×10−7 exp(−30[kJ mol−1]/RT) indicating a remarkable trapping effect on hydrogen diffusion caused by tiny oxide particles in the bulk of F82H.

Retention behaviors of tritium loaded near the surface region of metals by gas absorption and plasma implantation

Retention behaviors of hydrogen loaded by gas absorption and plasma implantation to pure copper, pure tungsten and the F82H steels at various temperatures have been examined by the tritium imaging plate technique. Three components are distinguished in hydrogen retained near the surface region; one is an endothermic trapping component in the bulk or near the surface region, second is an exothermic trapping component induced by plasma implantation and third is a trapping component in oxide layers. The relative amount of each component in depth near the surface region of the metals can alter retention behaviors of hydrogen with respect to the loading temperatures.

Study on kinetics of hydrogen dissolution and hydrogen solubility in oxides using imaging plate technique

Using a tritium imaging plate technique, kinetics of tritium dissolution and its solubility in several oxides were examined. Mirror-polished single crystals of alumina, spinel and zirconia were used as specimens, which were exposed to 133Pa of a tritium(T)–deuterium(D) gas mixture (T/(T+D)∼0.17) at temperatures ranging from 673 to 973K for 1–5h. The T surface activity on the specimens increased with increasing temperature and exposure time, it almost saturated at 873K and reached 2×105Bq/cm2 (1×1014T/cm2), and no clear difference appeared among the types of specimens. The T activity in the oxide bulk also increased with temperature, in which there was a trend for the oxides: spinel≧zirconia≧alumina. In the T dissolution process for all oxides, the concentration gradient due to its diffusion was not observed even for short exposure times: the T density was almost uniform over the specimens in transition states and increased with exposure time up to the saturated value. These experimental results suggested that the rate-controlling process of T dissolution in the temperature region should be not its diffusion in the oxides but dissociation of hydrogen molecules (T–D mixture in this case) into atoms, its adsorption on the surface and/or T penetration from the surface into the bulk.

Retention and desorption behavior of tritium in Si related ceramics

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 D2+ 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.

Depth profiling of hydrogen in ferritic/martensitic steels by means of a tritium imaging plate technique

Abstract In order to understand how hydrogen loaded by plasma in F82H is removed by annealing at elevated temperatures in vacuum, depth profiles of plasma-loaded hydrogen were examined by means of a tritium imaging plate technique. Owing to large hydrogen diffusion coefficients in F82H, the plasma-loaded hydrogen easily penetrates into a deeper region becoming solute hydrogen and desorbs by thermal annealing in vacuum. However the plasma-loading creates new hydrogen trapping sites having larger trapping energy than that for the intrinsic sites beyond the projected range of the loaded hydrogen. Some surface oxides also trap an appreciable amount of hydrogen which is more difficult to remove by the thermal annealing.

Retention and release mechanisms of tritium loaded in plasma-sprayed tungsten coatings by plasma exposure

Depth profiles of tritium (T) loaded by gas and plasma in tungsten (W) coatings on ferritic steels have been examined by using a tritium imaging plate technique and their changes during storage and after annealing have been monitored. The depth profiles of T consisted of 4 components, (I) T trapped at impurities and defects newly introduced in the near surface region of the coating by plasma loading, (II) T trapped at the inner surfaces of the grains and dissolved in the grains resulting in a flat depth profile throughout the whole coating, (III) T dissolved and diffused into the substrate giving a decaying profile, and (IV) T trapped at the backside surface of the substrate. The results support that retention of T is mainly caused by pore diffusion of gaseous T followed by dissolution and trapping in/at each W grain, and dissolution of T into the F82H substrate to allow permeation . Release of T proceeds in an opposite way of retention but each component desorbs independently.

Tritium Retention in First Wall Tile Gaps of JT-60U

Tritium (T) retentions in tile gaps (side surfaces) of the first wall of JT-60U were measured by a tritium imaging plate technique (TIPT). For all first wall tiles measured here, the T retention decreased from the front (entrance) to the bottom of the side surfaces showing superposing two exponential decays, which were already observed in the divertor region. Heavier erosion on the plasma-facing surface resulted in higher T retention in the front-side surfaces in the vicinity of the plasma-facing surface. In addition, wider gap width also resulted in higher T retention in the bottom side surfaces. Using the TIPT results, overall T retention in the side surfaces of the whole first wall was estimated to be ~6 × 1017 T atoms, which was only one-tenth of total T retention in the plasma-facing surface of the first wall in JT-60U.