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核融合施設では、各種化学形で漏洩しうるトリチウムを施設内に閉じ込める設備を設け、万一の事故の発生に対してもトリチウム除去設備を用いて影響を緩和させる。トリチウム閉じ込めとトリチウムモニタリング、核融合施設の異常事象を想定したトリチウム除去設備の設計と実証研究例を紹介するとともに、固相からのトリチウム回収のあり方につき解説する。

A detritiation system (DS) is required to remove tritium from the atmosphere of a nuclear contain-ment in any extraordinary situations. Realization of a DS that does not require heating of a catalyst reactor for tritium oxidation and frequent switching operation of adsorption columns for tritiated vapor collection will greatly contribute to the improvement of fusion safety. Concerning the catalyst reactor, it has been demonstrated that tritium can be oxidized at room temperature without any heating by the developed hydrophobic catalyst. To achieve a high tritium conversion efficiency for detritiation, it has already been revealed that suppression of production of tritiated hydrocarbons by hydrogenation reactions as side reactions of tritium oxidation in a catalyst reactor is the key issue to be solved. We have to pay special attention to ethylene among hydrocarbons because ethylene is easily tritiated by reaction of hydrogenation. In this study, complete combustion of ethylene at room temperature in the catalyst reactor is proposed as a measure to suppress the formation of tritiated hydrocarbons. Catalytic combustion characteristics of hydrocarbons were obtained, and the change in the ignition temperature by a change in each design parameter of the catalyst was demonstrated. Concerning noble metal species, platinum is superior to palladium due to less susceptibility to water vapor. The smaller the particle size of noble metal is, the higher the activity is, but because it is more susceptible to water vapor, the particle size of noble metal can be optimized. It was suggested that there is an optimum value for the pore size of the catalytic support.

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.

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.

In the fuel cycle of ITER and future fusion reactors, which consists of processes which are fueling, burning (by fusion reaction), vacuum pumping, purification (with recovery from tritiated impurities), isotope separation and storage, tritium accountancy control will be required from the viewpoint of operation and safety. Under ITER-EDA, as one of activities to establish the technological basis for tritium tracking and process system control, a remote and multi-location analysis system of hydrogen isotopes and tritiated molecular impurities in process gases by application of laser Raman spectroscopy was developed and validated at Tritium Process Laboratory (TPL) of JAEA [1]. Then, we had validated followability to process gas composition change and a detection limit of T2 gas of less than 1 kPa with an optimized system. More than 20 years after that development, great progress of laser and detection systems (especially, semiconductor lasers and Peltier-cooled CCD detectors, which have become very conventional and of high performance) has been achieved, which will enable us to make the system much simpler and conventional. A prototype system has been made at TPL, which consists of a semiconductor laser (488 nm, 200 mW, size: 31 mm×32.5 mm×63.5 mm), spectroscope (FL: 460 mm, 1200/mm grating) with a Peltier-cooled CCD detector, and an optical system. Results of some preliminary tests being conducted to confirm sensitivity (detection limits), stability, accuracy with static and dynamic test systems are presented., 3rd Asia Pacific Symposium on Tritium Science (APSOT-3)

核融合原型炉の一次燃料系において、プロトン導電体を用いた水素同位体移送によるダイレクトリサイクルを導入した場合の燃料系のインベントリーと応答への影響を評価した。ダイバータから水素同位体を選択的に気相で排気昇圧する結果、循環時定数が大幅に短縮され、インベントリー低減効果があることがわかった。, 日本原子力学会 2020年秋の大会

To understand the fuel retention mechanism, correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011-2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load.

Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign(2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles werea single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets,respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-materiallayers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thicknesswas ∼1.5 m. By means of transmission electron microscopy, nano-size bubble-like structures with asize of more than 100 nm were identified in that layer. They could be related to deuterium retention inthe layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition:a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction patternobtained with transmission electron microscope indicated Be was included in the layer. No bubble-likestructures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosionzone. This is consistent with the fact that the strike point was often located on that tile during the plasmaoperation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JETILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retentionand dust formation.

重水素-トリチウム燃料で動作する核融合炉は施設内に燃料循環システムを設ける。原型炉の燃料循環システムは、ブランケットで増殖されたトリチウムの燃料純度までの精製機能と高い信頼性を兼ね備え、前例のない大流量の重水素-トリチウム流を処理しなければならない。燃料循環システムの各機器の開発状況と原型炉に向けた燃料循環システムのあり方につき解説する。, NIFS一般共同研究 研究会「原型炉におけるトリチウムの課題」

The atmosphere detritiation system (DS) function is to remove tritium from the atmosphere to mitigate the impact on the environment in the case of accidental tritium release. The DS uses the technique of catalytic oxidation of tritium and collection of tritiated water vapor. The DS of ITER uses scrubber column technology for collection of tritiated water vapor, not water adsorption method conventionally used for DS, considering a failure occurrence probability of switching valves during long term operation. This technology is based on isotopic exchange between water vapor and liquid water to be arranged in a packed type column. This technology has never been applied to the DS of existing tritium facilities. Hence, qualification tests have been performed for detritiation to validate scrubber column technology for use at ITER with a pilot-scale scrubber column of 1/5 size of an ITER scale. The test column was filled with packing material and internal components of the same type as is used at ITER. In this qualification test, key objectives of the scrubber column design are to 1) collect tritium to meet DS design requirements, 2) achieve high reliability/availability and 3) avoid production of excessive tritiated water to be processed by WDS. The qualification test shows that detritiation factor (DF) is strongly influenced by a liquid flow rate with increasing the liquid flow rate resulting in higher DF. DF of scrubber column highly depends on lambda value which is defined as a flow rate of gas/liquid. A scrubber column of 2.88 m height with sufficient wettability of packing and efficient water dispersion by water distributor achieved high DF in the condition that lambda is one. Test results indicate that scrubber column technology achieves both high availability for detritiation and reduction of tritiated water production., 12th International Conference on Tritium Science and Technology (Tritium 2019)

Evaluation of hydrogen isotopes retention in the plasma-facing components (PFC)l is required for the design of a future fusion reactor. To obtain a set of relevant data on the isotope retention and their surface state (and chemistry), the analysis of PFC from the JET tokamak with the ITER-Like Wall (JET-ILW) is both useful and important. In previous studies, the correlation of surface chemical states and hydrogen isotope retention in the divertor samples from the first ILW campaign (ILW-1, 2011-2012, 19 h plasma operation, total energy 150 GJ) was determined. It was found that the major D desorption peaks were found at 643 K and 803 K. For samples at the strike point region, the D desorption temperature was clearly shifted towards higher T, indicating the stabilization of D trapping by higher heat load. In this work, the studied tiles were exposed during the 3rd campaign (ILW-3) which comprised in total 23 h of plasma with the total energy input of 245 GJ. The hydrogen isotope retention and chemical states of divertor tiles were studied by thermal desorption spectroscopy (TDS) and X-ray photoelectron spectroscopy (XPS). It was found that the major D desorption were located at 620 K and 800 K, which were consistent with the data for ILW-1. However, the total D retention was increased. The XPS analysis showed that major elements on the surface was Be with forming oxide state with co-existence of W. Carbon was found mainly on the outer divertor tiles. In the presentation, detailed XPS and TDS results will be shown and discuss the hydrogen isotope retention behavior in JET-ILW divertor tiles., 12 th International Conference on Tritium Science & Technology

Understandings of tritium inventory in plasma facing materials are important issues for safe maintenances and a control of fueling gasses in ITER and DEMO. Since dust particles are a floating substance, it is also important to consider a protection of radioactive material diffused from inside of the plasma vacuum vessel. A technical establishment for measuring tritium contained in dust particles was a long-term issue, and it was not resolved also in the case of D - T plasma operations in JET and TFTR. The ITER-Like Wall (JET-ILW) experiment as a real simulation of plasma facing materials towards ITER was started from 2011 in JET. The first campaign of the JET-ILW (ILW-1) operation 2011-2012 was followed by the retrieval of selected in-vessel materials for detailed ex-situ studies. Analyses of tritium amounts were done in QST, Rokkasho, Japan. At present, three kinds of methods, the combustion method for a quantitative total tritium measurement, the liquid scintillation counter method for simplified measurement of total tritium amount, and tritium imaging plate technique for tritium spatial distribution measurement, are available for dust particles. Specific tritium activities of dust particles in ILW-1 show similar amounts from a comparison with dust particles collected at the JET carbon wall campaign in 2007-2009. In the ILW-1 campaign, carbon dust particle remained, and TIP observation shows higher tritium inventory on carbon dust particles. In the 3rd ILW (ILW-3) campaign in 2014-2015, tritium amounts in dust particles were measured by the combustion method and specific tritium activity of 30-50 MBq/g was observed. This tritium amount in ILW-3 is one order of magnitude lower than the tritium amount about 700 MBq/g in ILW-1. From an energy dispersive X-ray spectroscopy (EDX) analysis on dust particles in ILW-3, number of carbon particles tends to decrease, compared with results in ILW-1, 12 th International Conference on Tritium Science & Technology

Detritiation system is required to remove tritium from the atmosphere of a nuclear containment in any extraordinary situations. Realization of detritiation system that does not require heating of catalyst reactor for tritium oxidation and frequent switching operation of adsorption columns for tritiated vapor collection will greatly contribute to the improvement of fusion safety. Concerning on the catalyst reactor, it has been demonstrated that tritium can be oxidized at room temperature without any heating by the developed hydrophobic catalyst. To achieve a high tritium conversion efficiency for detritiation, it has already revealed that suppression of production of tritiated hydrocarbons by hydrogenation reactions as side reactions of tritium oxidation in a catalyst reactor is the key issue to be solved. We have to pay special attention to ethylene among hydrocarbons because ethylene is easily tritiated by reaction of hydrogenation. In this study, complete combustion of ethylene at room temperature in the catalyst reactor is proposed as a measure to suppress the formation of tritiated hydrocarbons. Catalytic combustion characteristics of hydrocarbons were grasped and the change in the ignition temperature by change in each design parameter of catalyst was demonstrated. Concerning on noble metal species, platinum is superior to palladium due to less susceptibility to water vapor. The smaller the particle size of noble metal, the higher the activity, but because it is more susceptible to water vapor, the particle size of noble metal can be optimized. It was suggested that there is an optimum value for pore size of catalytic support., 23rd Topical Meeting on the Technology of Fusion Energy

Quantitative tritium inventory in dust particles from campaigns in the JET tokamak with the carbon wall (2007–2009) and the ITER-like wall (ILW 2011–2012) were determined by the liquid scintillation counter and the full combustion method. A feature of this full combustion method is that dust particles were covered by a tin (Sn) which reached 2100 K during combustion under oxygen flow. The specific tritium inventory for samples from JET with carbon and with metal walls was measured and found to be similar. However, the total tritium inventory in dust particles from the ILW experiment was significantly smaller in comparison to the carbon wall due to the lower amount of dust particles generated in the presence of metal walls.

We have applied a tritium imaging plate technique (TIPT) in combination with an electron-probe microscopic analysis (EPMA) to examine a tritium (T) retention characteristic in individual dust particle collected in the Joint European Torus with the ITER-like Wall (JET-ILW) after the first campaign in 2011-2012. Dust particles were collected from the inner divertor region in JET-ILW. A tiny amount of dust was taken from that sample and placed on the surface of a disk made of indium (In). The content of T retained at a surface of and/or in the dust particles was evaluated by TIPT: emission of beta-rays from T to the imaging plate (IP) was measured over a whole surface of the indium disk. Afterwards, EPMA was conducted to determine morphology (composition and structure) of the dust particles in the same surface area where the IP measurement was conducted. The obtained maps (images) of the T distribution and elemental compositions were super-imposed to reveal characteristic features of T retention in the individual dust particle. The results suggest that most of T was retained at the surface of and/or in carbon (C)-dominated dust particles. The characteristics of T retention in C-dominated dust particles are scattered. The retention in tungsten, beryllium and other metal-dominated dust particles is relatively small: smaller by a factor of 10 ~ 100 in comparison with that in the C-dominated particles., The 23rd International Conference on Plasma Surface Interactions in Controlled Fusion Devices (PSI-23)

Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points: (a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1–1.4 g per campaign (19.1–23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JETC). (b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2–10 μm in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters. (c) Tungsten dust occurs mainly as partly molten flakes originating from the W-coated tiles., The 13the International Symposium on Fusion Nuclear Technology(ISFNT-13)

Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified (“geological-like”) mixed-material deposition layer which mainly included Be and Ni with the thickness of ∼2 μm. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper., The 13the International Symposium on Fusion Nuclear Technology(ISFNT-13)

Comprehensive analysis of the divertor tile surface and dust produced by the 1st campaign (2011-2012) of the JET ITER-Like-Wall (ILW) experiment has been performed in QST Rokkasho as a part of the Broader Approach (BA) DEMO R&D activity. The R&D facility is a unique facility where tritium, beryllium and activation materials can be handled. Plasma-Wall-Interaction analyses on these samples were carried out by NIFS, universities and QST collaboration with strong support by JET and EU. Microstructure, atomic composition and chemical states of the surfaces on the tiles and dust have been analysed. The divertor target in JET-ILW is made partly of tungsten (W) bulk lamellar plates and mainly of W-coating on the carbon fiber composite (CFC) bulk tiles by Vacuum Plasma Spray (VPS) process. In this analysis, W-coating tile samples (diameter of 17mm) were used, retracted from the inboard and outboard vertical targets and at the divertor floor. Two bottles of dust samples were collected by a vacuum cleaner on the inner and oud outer divertor surfaces. In JET-ILW, most of the first wall surface is covered by Be, and Be impurity is transported with the plasma from the first wall to the divertor with repeating the deposition and erosion processes. Recently, some unique analysis results were obtained, in addition to standard surface analysis. Microstructure analysis of tile and dust cross-sections as well as their surfaces have been performed, using a transmission electron microscope (TEM) after the focused ion beam (FIB) fabrication. In particular, in the thick Be-deposition layers on the inner W-target, nano-size bubbles-like structure was generated, and the deposition layers were composed by mixture of W, Be, C, Ni and O, evaluated by using X-ray photoelectron spectroscopy (XPS) and Electron Probe Micro Analyzer (EPMA). These deposition layers might be affected by various plasma operations in the experiment campaign. Local pattern of the Be-deposition on a rough VPS-W surface against the plasma incident angle was investigated. The micro-structure analysis of dust particles showed that bubble-like structure was also generated inside a Be dominant dust (large 100 μm size dusts were so far analyzed), and chemical components and state analysis showed that BeO in a Be dominant dusts and some W and/or C dusts were determined. The analysis method of the tile and dust samples were established. Evaluations of hydrogen isotope retention and Tritium-specific activity (Bq/g) in JET-ILW tile surface and dust particles will be also presented., 13th International Workshop on Beryllium Technology

To understand the fuel retention mechanism, correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011-2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load., 2nd Asia-Pacific Symposium on Tritium science

Tritium (T) distributions in divertor tiles retrieved from the JET-ITER Like Wall (ILW) after the 1st (2011–2012) and 2nd (2013–2014) experimental campaigns were examined by measuring energy spectra of X-rays induced by β-rays from T. Disk specimens cored from the W-coated CFC divertor tiles (Tiles 1, 4, 6, 7 and 8) were analysed using a silicon-drift detector (SDD) equipped with an ultra-thin window (70 nm-thick Si3N4) at QST, Rokkasho. Analysis using a SDD with a conventional window (8 μm-thick Be) was performed at VTT. The former SDD was used to detect characteristic X-rays of Be together with those of W, while the latter was more suitable for non-destructive depth profiling by analysing X-ray energy spectra. A majority of T currently present in the JET vacuum vessel is produced by DD fusion reactions [D + D → T (1.01 MeV) + H (3.03 MeV)]. Therefore, T is present as a minor isotope of hydrogen in the vessel. A part of T produced is implanted into the tiles at high energy (~MeV), while a remaining part of T is thermalized in the plasma. Strong peaks of Be KαX-rays were observed for the horizontal and upper vertical parts of Tile 1 due to heavy co-deposition of T with Be eroded from the main chamber wall. Intensity of Be KαX-rays was below detection limit for other tiles because of far smaller thickness or absence of Be deposition layer. In contrast, the intensity of W MαX-rays was almost uniform though it was somewhat high at the positions close to the main plasma (the horizontal parts of Tiles 1 and 8). Depth profile analyses showed that T penetrated into depths up to 0.5–4 μm. These observations indicated the implantation of high energy T into the W layers. Exceptionally high intensity of X-rays (and also γ-rays) was observed at the shadowed region of Tile 4, though thermal desorption spectrometry showed relatively small D retention in this region. The mechanisms underlying this exceptional T accumulation will be discussed in the presentation by taking account of the results of microstructural analysis., 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC-16)

A comprehensive analysis of collected dust particles in the Joint European Torus ITER-like Wall (JET-ILW) after the first campaign in 2011-2012[1] has been carried out at the International Fusion Energy Research Centre (IFERC) in order to identify dust characteristics such as structures, material components and hydrogen isotope retention. Analysis of the cross-section structure as well as surface one began with from a large-size dust particle (~100 m). For the flake-type beryllium (Be)-based dust, it was found that the damaged Be crystal structure contained a larger oxygen component near the surface (~2 m). The specific dust flake of beryllium oxide in the experiment device was shown for the first time. Deuterium retention in small weight of dust particles (4.4 mg) was evaluated to be 1.2 x 1021 atoms/g by TDS, which estimated to be 8.2 x 1020 atoms for entire dust particles collected from the inner divertor (0.7 g). This result firstly suggests that a contribution of dust particles to the total retention in the experiment was small, i.e. less than 1% of the total retention in deposition layers of the inner divertor target., 26th IAEA Fusion Energy Conference

Comprehensive ex-situ analyses for JET dust particles were started in Rokkasho center, and it has been carried out based on Broader Approach (BA) agreement. Analyses for ILW dust particles at inner divertor were started from June, 2015. Surface morphologies for ILW dust particles are measured by SEM/EDX. Typical diameter of dust particles is less than 100 micro-m. Be flake dust (thickness of 8 micro-m) showed oxide layers of 2 micro-m. Tritium retention of ILW dust particles at the inner divertor was evaluated to be 0.1GBq/g by LSC. Deuterium retention of ILW dust particles was evaluated to be 1.2 x 1021 atoms/g from the total dust weight of 4.4mg at the inner divertor (TDS)., 22nd International Conference on Plasma Surface Interactions in Controlled Fusion Devices