Tritium Imaging Plate Technique Research Articles

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.

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.

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.

Application of a Tritium Imaging Plate Technique to Depth Profiling of Hydrogen in Metals and Determination of Hydrogen Diffusion Coefficients

Application of a Tritium Imaging Plate Technique to Depth Profiling of Hydrogen in Metals and Determination of Hydrogen Diffusion Coefficients

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.

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.

Behaviour of tritium in plasma-sprayed tungsten coating on steel exposed to tritium plasma

In order to understand the role of a plasma-sprayed tungsten (W) coating on tritium (T) permeation in a W-coated ferritic/martensitic steel (F82H), we have examined depth profiles of T in the coating and the substrate using the tritium imaging plate technique after T loading by a dc glow-discharged plasma at 453 and 573 K for 2 h. Tritium loaded by plasma exposure was distributed uniformly in the whole coating, while T penetrated to the substrate by diffusion. The former is caused by T diffusion through open pores and/or along grain boundaries followed by adsorption on grain surfaces and dissolution in the grains. The main role of the W coating on T permeation is to reduce the incoming flux at the coating/substrate interface owing to pore diffusion in the coating and the effective area for T dissolution in the substrate.

Behavior of Tritium near Surface Region of Metals Exposed to Tritium Plasma

In order to understand behavior of tritium (T) on surface and in bulk of metals exposed to T plasma, both surface activities and depth profiles of T were periodically observed by a tritium imaging plate technique during storage in air at room temperature (RT) for over 1 year. In the T depth profiles, T localized within a depth of sub mm from the surface was clearly distinguished from T in the bulk. The former was attributed to strong trapping by some defects produced by the plasma exposure and remained quite longer during the storage, while the latter was released from the surfaces by diffusion. T surface activity measured on the plasma-exposed surface changed in a complicated way with time due to removal of T by isotopic replacement with H in ubiquitous H2O and T supply from the bulk in the course of the diffusional release.

Tritium Distribution on First Wall Carbon Tiles in JT-60U

T retention and its depth profile in the graphite tiles used for first wall of JT-60U have been measured by a tritium imaging plate technique and a full combustion method. T was found only limited depth beneath the plasma facing surface and little in both the surface region shallow than 1 μm and in bulk more than 1mm in depth. Although most of T produced by DD reactions are thermalized and neutralized in plasma and impinge on the plasma facing surface and penetrate into the inner surface, they are isotopically replaced by subsequently incoming D. Only some of high energy T escaping from plasma are directly implanted beneath the surface and retained escaping from the isotopic replacement until attainment of a saturation concentration.

Application of tritium tracer techniques to observation of hydrogen on surface and in bulk of F82H

Hydrogen including a trace amount of tritium was loaded on the edge surface of an F82H rod. After the loading, the rod was held at 298 or 323K to allow hydrogen diffuse in and release out. Tritium tracer techniques have been applied to determine hydrogen depth profiles and hydrogen release rates by using an tritium imaging plate technique and a liquid scintillation counting technique, respectively. The depth profiles were composed of a surface localized component within 200μm of the surface and a diffused component extending over 1mm in depth. The apparent hydrogen diffusion coefficients obtained from the depth profile of the diffused component are near the extrapolated value of the literature data determined at higher temperatures. The surface localized component, which is attributed to trapping at surface oxides and/or defects, was released very slowly to give apparent diffusion coefficients much smaller than those determined from the diffused component.