Imaging Plate Technique Research Articles

3D benchmark experiments of tritium in tungsten for tritium measurements by detecting β-ray induced X-rays

Tungsten is one of the most promising materials for plasma facing components (PFCs), and β-ray induced X-ray spectrometry (BIXS) is an important non-destructive method to obtain tritium depth profile and retention information for the development of PFCs, recovery of fuel and control of tritium safety in a fusion reactor. In this article, benchmark experiments to obtain tritium 3D profile in tungsten have been firstly performed using layer-by-layer chemical etching (LLCE) and imaging plate (IP) techniques. BIXS spectra have been detected at each layer. The average layer thickness of each erosion was controlled to be around 150 nm and 8 layers (the total thickness was 1.21 μm) were eroded, and the maximum uncertainty of tritium activity was 7.91 % in the LLCE experiments. Monte-Carlo simulations of BIXS spectra with Geant 4 code and four kinds of physical packages (Penelope, Livermore, EmStd. Opt1 and EmStd. Opt4) have been benchmarked against to the experimental results. Results show that all the four physical packages are not suitable to calculate the intensity of the BIXS spectra, and the current mode or cross sections should be improved to simulate the interaction between low electron/photon and tungsten accurately. However, the relative intensity of W-Mα peak and W-Lα can be simulated well using current physical packages such as EmStd. Opt4 and Livermore, which is a good indicator for tritium depth profile in tungsten.

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.

An approach for 3-D distribution measurement of tritium in tungsten by chemical etching and imaging plate technique

Measurement of tritium distribution details both on surface and in bulk of tungsten is of great importance for studies on the retention properties of plasma facing materials (PFMs) in magnetic fusion reactors. An attempt was made to determine the tritium 3-D distribution in tungsten sample through layer by layer chemical etching method (LLCE) and imaging plate technique (IP). As chemical etching method has been rarely employed for tritium measurement in tungsten, the influence of key experimental parameters were evaluated. And the tritium 3-D distribution in a tungsten sample was obtained by LLCE and IP. It is found that LLCE is a promising and reliable method for tritium depth profile measurement. The relative error of tritium content in each layer is approximately 5.2% when the thickness of erosion layer is 0.4 μm. Combined with IP technique, the detailed 3-D distribution could be obtained and it is clearly indicated that tritium is unevenly distributed on the radial section of this sample. In addition, the good agreement of tritium depth profile by LLCE and IP shows the reliability of this approach.

Global distribution of tritium in JET with the ITER-like wall

• Tritium (T) distributions in Be, W and W-coated CFC tiles from JET were examined. • The center of Be outer limiter showed the highest T retention in the main chamber. • The co-deposits in the divertor with high C fraction showed high T content. • The C fraction and T content after 3rd campaign were low compared to 1st campaign. • T retention in the bulk W tile was far smaller than that in the W-coated CFC tiles. Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed.

Investigation of the distribution of remaining tritium in divertor in LHD

Abstract To reveal the triton transport and the tritium migration in a deuterium plasma experiment in the Large Helical Device (LHD), the distribution of the remaining tritium in divertor tiles made of graphite after the first deuterium plasma experimental campaign in 2017 was investigated. In this study, tritium contents in divertor tiles have been measured by using a full-combustion method. The asymmetric tritium retention in divertor tiles located at symmetric positions, which was found in the previous study by the surface tritium measurement using an imaging plate technique, has been confirmed by the results of the full-combustion method. The asymmetry is considered to be attributed to the asymmetric distribution of lost-points of energetic tritons in divertor. A depth profile of remaining tritium in a divertor tile estimated by using a combination of the imaging plate technique and a sputtering treatment shows that the peak of the profile locates at several micro-meters from the surface. This result suggests that the majority of the remaining tritium impinged upon the divertor tile as energetic tritons. In this study, a distribution of lost-points of energetic tritons has been calculated by using a Lorentz orbit following code (LORBIT) with taking into account divertor components. The obtained distribution has been compared with measured tritium distributions on divertor tiles. The measured and calculated distributions are similar to each other, but they are not the same. The difference between them can be attributed to plasma exposures during and after the deuterium plasma campaign.

Tritium distribution analysis of Be limiter tiles from JET-ITER like wall campaigns using imaging plate technique and β-ray induced X-ray spectrometry

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.

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.

Visualization of hydrogen isotope distribution in yttrium and cobalt doped barium zirconates

Tritium (T) distributions in yttrium (Y) and cobalt (Co) doped barium-zirconates (BaZrO3) were visualized by imaging plate (IP) technique using HT gas (T/H = 10−4, ∼1.3 kPa) under various exposure temperatures (623 to 1273 K) and exposure times (5 min to 16 h). Two types of specimens, BaZr0.9Y0.1O3-α (BZY) and BaZr0.955Y0.03Co0.015O3-α (BZYC), were prepared with conventional powder metallurgy: disk-shaped specimens were obtained with more than 97% of the theoretical densities. With IP method, uniform tritium distribution was found on the surface of both T-exposed specimens. The T distributions of the cross sections of the cut-in-half specimens were also examined with IP, which showed that more T was diffused into BZYC compared to BZY. The diffusivities of T and hydrogen solubility were determined by the T concentration profiles for the BZY and BZYC samples at different exposure conditions. Almost one order higher solubility and diffusivity for BZYC than BZY was found. It suggests that only a small amount sintering-aid cobalt doping along with yttrium may play a vital role of enhancement of electrochemical activity of the proton conducting BaZrO3 under HT exposure condition.

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.

Neutron flux distributions in the Large Helical Device torus hall evaluated by an imaging plate technique in the first campaign of the deuterium plasma experiment

The global flux distributions for thermal, epithermal and fast neutrons in the torus hall of large fusion devices were experimentally evaluated for the first time in the Large Helical Device (LHD) using the activation foil method measured by the imaging plate and high-purity germanium detector. It turned out that the thermal neutrons were effectively absorbed by borated polyethylene blocks placed beneath the LHD. This should reduce the radioactivity of the floor and would be beneficial in maintaining a good environment for radiation workers. Uniform distributions of epithermal and fast neutrons were observed near the LHD. In particular, the significant decrease in fast neutron flux with increasing distance from the LHD, due to the fast energy loss of fast neutrons, was observed.The neutron flux distribution measurement, with rough energy discrimination based on the threshold energy of the neutron activation foil, allows us to estimate the spatial radiation dose rate, as well as the radioactivity in components in the torus hall. The prediction of the radioactivity in the concrete floor indicated that radioactive isotope 55Fe will be a dominant source of radioactivity in the concrete after the nine-year deuterium experiment campaign finishes.

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.

Electronic autoradiography of 133Ba particle emissions; diffusion profiles in granitic rocks

The spatial distribution of barium activity in granitic rocks was measured with two autoradiography techniques; digital autoradiography using phosphor imaging plate technique (Fuji 5100) and filmless electronic autoradiography (i.e. The BeaQuant™), which is based on a gas detector incorporated in a micromesh Parallel Ionization Multiplier (PIM). Rock cubes taken from a diffusion experiment that were in contact with 133Ba tracer were measured to determine diffusion profiles. In addition, the spatial distribution of 133Ba in the samples was determined. Polymethyl methacrylate standards for 133Ba were developed to determine the counting efficiency for electronic autoradiography. Good visual correlation between the two autoradiography methods were obtained in this study. The results of the experiments presented here can be utilized in future studies on the diffusion behavior of barium in granitic rocks.

Tritium distributions on W-coated divertor tiles used in the third JET ITER-like wall campaign

Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015–2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and β-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013–2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign.

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.

Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011–2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

Measurement of tritium in tungsten deposition layer by imaging plate technique after exposure to gaseous tritium

It is important to understand tritium desorption behavior from plasma-facing materials of a fusion reactor in order to discuss effective tritium recovery method from in-vessel components. However, basic behavior of hydrogen isotopes in W deposition layer is not understood completely. In this study, characterized tungsten deposition layer formed by hydrogen plasma sputtering was exposed to gaseous tritium at 300°C or 500°C and tritium desorption behavior by vacuum heating was investigated by the imaging plate technique. For comparison, bare tungsten substrates were exposed to gaseous tritium in the same condition. Initial tritium activity in the deposition layer was much higher than that in the bare substrate. Tritium desorption behavior from tungsten deposition layer was different by the temperature of the layer during tritium exposure process. By heating at 500°C for 1h, 97.5% of tritium was desorbed from the layer exposed to tritium at 300°C. On the other hand, by heating at 500°C for 2h, only 44.6% of tritium was desorbed from the layer exposed to tritium at 500°C. To recover most tritium from W deposition layer and W substrate, heating at above 700°C is required.

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 distributions on tungsten and carbon tiles used in the JET divertor

Tritium distributions on the W-coated divertor tiles used with Be wall in JET 2011–2012 ITER-like wall (JET-ILW) campaign were measured using an imaging plate (IP) technique. The high intensity of photo-stimulated luminescence (PSL) from IP was observed at the regions covered by deposited Be layers. However, the PSL intensity was not simply proportional to the thickness of the deposited Be layers; the shadowed region of Tile 4 showed the highest PSL intensity though the thickness of deposited Be layer on this region was smaller than that on Tile 0 and the apron of Tile 1 by an order of magnitude. These observations indicated the influence of impurities such as oxygen on tritium retention in the deposited Be layers. The C tiles used in the 2007–2009 JET carbon wall (JET-C) campaign were also examined. The high PSL intensity was observed for the regions covered with deposited C layers in this case. The area of tile surfaces covered by the deposited tritium-rich layers on the W-coated-tiles used in the JET-ILW campaign was significantly smaller than that on the C tiles used in the JET-C campaign.