Tritium in plasma-facing components of JET with the ITER-Like-Wall

The ITER-Like-Wall project has been carried out at the Joint European Torus (JET) to test plasma facing materials relevant to ITER. Materials being tested include both bulk metals (Be and W) and coatings. Tritium accumulation mechanisms and release properties depend both on the wall components, their location in the vacuum vessel, conditions of exposure to plasma and to the material itself. In this study, bulk beryllium limiter tiles, plasma-facing beryllium coated Inconel components from the main chamber, bulk tungsten and tungsten coated carbon fibre composite divertor tiles were analysed. A range of methods have been developed and applied in order to obtain a comprehensive overview on tritium retention and behaviour in different materials of plasma facing components (PFCs). Tritium content and chemical state were studied by the means of chemical or electrochemical dissolution methods and thermal desorption spectroscopy. Tritium distribution in the vacuum vessel and factors affecting its accumulation have been assessed and discussed.


Introduction
Tokamak type reactors are the dominant experimental technique for studying nuclear fusion. One of the largest present-day tokamak device, The Joint European Torus (JET), located near Oxford in United Kingdom, was put into operation in 1983. Currently, the International Thermonuclear Experimental Reactor ITER is under construction in Cadarache, Southern France. ITER-Like-Wall project has been carried out at the JET in order to study the compatibility of plasma operation with the plasma facing materials relevant to ITER and estimate fuel retention and material migration [1]. Since 2011, three series of campaigns had taken place: ILW1 (2011-2012), ILW2 (2013-2014) and ILW3 (2015)(2016). During the shutdowns wall components had been retrieved for ex situ measurements. In this study, tritium concentration in different materials of the vacuum vessel has been measured, tritium accumulation pattern in the samples exposed for the separate ILW campaigns and for all the period assessed.

Plasma facing components in JET ITER like wall
Materials being tested include bulk beryllium, bulk tungsten, beryllium coated Inconel and tungsten coated carbon fibre composite. Positions of each material type in the vacuum vessel are illustrated in figure 1.
Plasma facing materials possess very different properties and, therefore, each requires a different approach for the measurements of tritium trapped in the surface, sub-surface and the bulk.
Beryllium's advantages as a plasma facing material are its low Z, good thermal conductivity, and high oxygen gettering characteristics. It has been tested as a plasma facing material also in earlier experiments both in JET, with its first introduction in 1989, as well as a smaller early period tokamaks such as UNITOR and ISX-B [2]. The main drawback of the use of beryllium is its high toxicity. A comprehensive overview on beryllium as PFM is provided by Temmerman et al [3]. Any further distribution of this work must maintain attribution to the author(s) and the title of the work, journal citation and DOI.
Tungsten has good thermal properties with a very high melting point, however, it can cause deleterious radiation losses if entering the plasma. Therefore, tungsten and its coatings are used only in the vessel location where high thermal loads are expected.
The use of carbon based materials has a long history since development of nuclear fission reactors -carbon has very good thermal properties, high melting point and low Z [4]. However, there is a limitation due to high fuel retention that has been observed during full carbon wall campaigns at JET [5]. Fuel retention in CFC is related to intense dust generation [5] and hydrogen chemical affinity to carbon [6]; both lead to the intense tritium co-deposition.
Inconel is a nickel-based alloy with following chemical composition: Ni 53 Fe 19 Cr 19 NbMoTi. It has been chosen as a vessel construction material due to its good oxidation and corrosion resistance, as well as good mechanical and thermal properties. It is also used as a substrate for beryllium coating due to the matching thermal expansion coefficients (11.4·10 −6 K −1 and 11.2·10 −6 K −1 for beryllium and Inconel, respectively).

Tritium sources in JET
Up to 2021, no tritium introduction was done within the ITER-Like-Wall campaign. Nevertheless, there are several other possible sources of tritium: • in-vessel tritium inventory remaining from previous D-T campaigns; • energetic tritium ion production as a result of D-D reaction; • tritium production in neutron-induced transmutation of beryllium (can be considered negligible in these results as we report in the following).
Tritium has been introduced in the vacuum vessel in three campaigns: Preliminary Tritium Experiment (PTE) in 1991 [7] the first Deuterium-Tritium Experiment (DTE1) in 1997 [8,9] and the Trace Tritium Experiment (TTE) in 2003 [10]. In total 35.39 g (5 ·10 24 atoms or 9 PBq) of tritium have been introduced (5 mg -PTE, 35 g-DTE1 and 380 mg-TTE) [11]. The highest contribution was the DTE1 experiment with the largest tritium amount being introduced into the torus, mainly by gas puffing. DTE1 was followed by clean-up period of about 3 months in order to remove tritium by running discharges in H, D or Glow Discharge cleaning mode, baking and venting [5]. Moreover, most of remained tritium must be removed together with the extracted CFC tiles during the exchange to metallic wall components. Additionally, tritium decay also must be taken into account as its half-life is comparably short − 12.32 years [12].
Tritium produced as a result of D-D reaction (1) has its own kinetic energy of 1.01 MeV [11] that leads to its ability to implant into the material.
Both gas puffed and D-D tritium ion behaviour in plasma has been studied in Tokamak Fusion Test Reactor TFTR [13] where both D-D and D-T fusion reactions have been achieved [14,15]. Neutron induced transmutation of beryllium will have a significant role in the future fusion devices such as ITER. Beryllium neutron interactions have been studied widely, especially, regarding its planned use as a neutron multiplier in tritium breeding blanket [16][17][18]. An example of Be transmutation chain where tritium can be produced is given in the following reactions 2-4 [19]. Overall neutron yield in ILW campaign was very low and its contribution to tritium generation can be considered as insignificant. According to Fonnesu et al [20], neutron budget in the ILW1, ILW2 and ILW3 was 7.71·10 18 , 1.85·10 18 and 2.15·10 19 , respectively (or ∼3.5-10·10 16 m −2 if vessel inner wall surface of 220 m 2 is used in calculations [21]). According to the data from fast neutron (> 1MeV) irradiation experiments of beryllium at High Flux Reactor in Petten [22], irradiation to neutron flux up to 2·10 26 m −2 generates about 640 appm or 4.3·10 19 atoms of tritium per gram of beryllium [18]. In case of neutron fluxes at JET, concentration of tritium would be ∼10 10 atoms that is near or below detection limit of measurement devices.

Tritium interaction with plasma facing materials, retention mechanisms
Co-deposition with eroded materials and physical diffusion into the bulk of wall material are being considered as the main tritium accumulation mechanisms in the first wall materials. Co-deposited tritium mainly retains in near surface of the plasma facing material, however tritium as a hydrogen isotope is very mobile due to its small size and might diffuse much deeper into the bulk. Energetic ion implantation into the material has been also widely studied [23].

Tritium measurement methods
There are various approaches for tritium measurements depending on the material and required information. Thermal Desorption spectrometry TDS is a standard method providing information on total amount of tritium, estimations of detrapping energies and diffusion is also possible. TDS can be performed both in a purge gas and vacuum, tritium measurement-by the means of mass spectrometry or radiometric methods. TDS has been applied both for deuterium and tritium measurements [24][25][26], however, tritium concentration in JET samples is very low and high sensitivity detection method is required.
The full combustion method followed by tritium measurements with liquid scintillation can be applied to all combustible materials, such as carbon and metals. This method has been originally developed by Vance et al [27] and has been applied to various JET materials [28,29].
X-ray photoelectron spectrometry [30] has a high resolution (nm), however it is limited only to near surface tritium measurement: 3 μm and 0.3 μm for Be and W, respectively. Method has been applied to the JET vacuum vessel tiles retrieved during ILW shutdowns [29,31].
Imaging plate technique is based on tritium beta radiation interaction with radiosensitive material. Due to small beta radiation range in solid samples method gives information only on the near surface tritium [32].
Glow discharge optical emission spectroscopy (GD-OES) is based on simultaneous surface etching with Ar plasma and optical stimulation of the eroded material. Currently, this method has been demonstrated on stable hydrogen isotopes H and D, only [33].
Chemical and electrochemical etching can be applied to the metallic samples and gives information both on total tritium amount and its depth profile [34] 2. Experimental

Materials
In this study, tiles retrieved during ILW1, ILW2 and ILW3 shutdowns have been analysed.
Those include bulk beryllium tiles from inner, outer walls and upper part of the vessel, beryllium coated Inconel from inner wall and bulk tungsten and tungsten coated CFC tiles from three divertor positions. Some beryllium samples have been in the vacuum vessel throughout all 3 campaigns. Locations of the studied samples and their photographs are provided in figure 2.

Methods
Dissolution/etching method is based on the simultaneous etching of the metal surface and measurements of the released tritium. Beryllium can be etched chemically in acid solutions, whereas for tungsten and Inconel electrochemistry must be applied due either to their low reactivity with chemical etchants or formation of complex mixtures of the reaction products. Detailed description of the methods for Be and W is given in [34]. In case of Be coated Inconel combination of chemical (beryllium) and electrochemical (Inconel) etching has been applied that gives opportunity to measure separately tritium accumulated in the coating and substrate. Released tritium in Ar purge gas (T 2 , HT, DT) has been measured by proportional counter with an operating volume of 300 cm 3 and a tritium monitor TEM 2102A (Mab Solutions GmbH), whereas tritium remaining in solution (HTO, DTO, T + ) with liquid scintillation counter TRi-Carb 2910TR (PerkinElmer, Inc.).
Full combustion method has been applied to tungsten coated CFC samples. Released tritium oxidized into water (HTO, DTO) and measured by liquid scintillation.
Thermal desorption of tritium was performed in a flow of He+0.1% H 2 purge gas. A sample is placed and heated in a quartz tube with two compartments-one for the sample and one for a bed of granulated zinc. A quartz cap of the tube had a thermocouple channel. The zinc bed is used to convert tritiated water to molecular gaseous tritium (HT, DT, and traces of T 2 ). The temperatures of the sample, the zinc bed and the cold trap were continuously measured. The tritium activity in the purge gas was continuously monitored using a proportional counter (similar model as described above). Heating rate of 5 K has been used with a maximum temperature 1305 K.
Tritium decay has been considered and all the measurement data recalculated as per day of the retrieval of the material from the vacuum vessel.

Results and discussion
Total amount of tritium measured by different methods, its surface and bulk distribution, retention behaviour in different plasma facing materials exposed to D-D plasma during ILW have been discussed in this manuscript.   1, figure 3). Tritium concentration has been found to be in range from 1·10 12 to 4·10 13 atoms cm −2 with its highest concentration in outer wall bulk beryllium limiter tiles. In contrast to deuterium data, no increase of tritium concentration was observed in the deposition zones of divertor, where its concentration was in range 10 12 to 10 13 atoms/cm 2 . According to Widdowson et al [35] deuterium concentration in inner divertor and inner wall limiter tiles ratio were about 10:1 (17·10 22 and 1.4·10 22 D atoms, respectively), whereas for tritium it is∼1:1 (5·10 12 and·10 12 T atoms, respectively). These results indicate that co-deposition is not the main tritium source in wall materials. A detailed comparison of tritium data with deuterium in bulk beryllium tiles and differences of ILW series has been published elsewhere [36].  When comparing tritium accumulation in samples with similar toroidal locations (figure 4) from each campaign with those that remained in vacuum vessel throughout all 3 campaigns it was observed that accumulated tritium final amount is smaller than a sum of three separate campaigns.
These results demonstrates the simultaneous desorption process taking place in the vacuum vessel, mostly due to bake-outs, outgassing and erosion [37]. However, in separate samples exposed to all campaigns tritium concentration is higher that expected form the sum of the separate campaigns. That might be explained by the fact that samples are taken from slightly different poloidal positions. Comparison of the tritium concentration values recalculated to ILW3 retrieval day taking into account decay of tritium is given in the table 2.

Effects of deposition layer on tritium retention
The impact of the deposition layer thickness on tritium retention pattern has been assessed. Combination of scanning electron microscopy and energy dispersive x-ray spectroscopy was used for the layer analysis ( Figure 5).  For the determination of the carbon deposit layer, SEM/EDX line spectrum has been acquired and measured. In figure 5 -line spectrum of the beryllium sample cross section. Three distinct areas can be identified-deposition layer mainly consisting of carbon at the plasma facing, beryllium roughening and oxidation in the middle and intact beryllium deeper in the bulk. Thickness of the deposition layer has been measured for the selected samples prior to tritium measurement and correlation assessed ( Table 3).   A linear correlation can be seen when looking at acquired data in logarithmic scale (R 2 =0.908)-with increasing deposition layer the retained tritium concentration decreases ( figure 6). A different trend can be observed for the samples retrieved from vacuum vessel than the ones acquired from artificial samples where tritium is co-deposited with carbon [38]. Multiple studies about the increase of deuterium retention by increasing the deposition layer can be found [38,39] and many tend to assume similar trends for tritium, but current results show the opposite tendencies. However, these results suggest that high energy T implantation is dominant for tritium retention at the plasma facing surface instead of co-deposition as it is in the case of deuterium.
This is due to the D-D reaction during deuterium campaigns, in which high energy tritium ions are created. In further planned JET deuterium-tritium campaign and in ITER the dominant reaction will be D-T. However, also D-D reaction will be present, therefore high energy tritium ions must be considered. Deposits may have larger impact on tritium retention at grooves [40].
For tritium, main concerns associated with deposition layer formation are complicated desorption of tritium already accumulated. ITER bake-out temperature is planned to be 510 K, which might be enough for beryllium materials, but the desorbed deuterium and tritium might still be retained in the carbon deposit layer [41,42].

Chemical state of tritium accumulated in Be
Dissolution method provides possibility to acquire information about possible chemical state of tritium [43]. During dissolution process of Be in sulphuric acid tritium trapped as interstitial atom or as a gaseous molecule (HT, DT or T 2 ) is released in gas phase and measured in the purge gas by the means of tritium monitor, whereas chemically bonded tritium remains in the liquid and is measured by liquid scintillation. Chemical reactions taking place during dissolution are as follows: