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Long term fuel retention experiments have been performed in JET with the ITER- Like Wall (JET-ILW) and compared with reference discharges in the preceding phase with carbon wall (JET-C) in ohmic, L and H-mode plasma scenarios. The long term fuel retention is evaluated through global gas balance with an accuracy of 1.2% for series of repetitive pulses (10–25) carried out over a full day of experiments. Compared to carbon wall, with the JET-ILW, for L mode, the retention exhibits also a drop from 1.26 × 1021 Ds−1 to 4–8 × 1019 Ds−1. Finally for Type III and type I ELMy H-mode, the retention decreases from 1.37 × 1021 Ds−1 to 7.2 × 1019 Ds−1 and from 1.7 × 1021 Ds−1 to 2.7 × 1020 Ds−1 respectively. The retention rates with the JET-ILW exhibit a decrease by a factor of about 10 compared to JET-C attributed to a reduction of carbon impurities and less fuel content in Be codeposition.

The residual carbon content and carbon edge flux in JET have been assessed by three independent diagnostic techniques after start of plasma operation with the ITER-Like Wall (ILW) with beryllium first wall and tungsten divertor: (i) in-situ measurements with optical spectroscopy on low ionisation stages of carbon, (ii) charge-exchange recombination spectroscopy, and (iii) residual gas composition analysis in dedicated global gas balance experiments. Direct comparison experiments in L-mode discharges were carried out between references from the previously installed material configuration with plasma-facing components made of carbon-fibre composite (JET-CFC) and the JET-ILW. The temporal evolution of the C divertor flux since installation of the İLW has been studied in the ohmic phase of dedicated monitoring discharges which have been executed regularly throughout the experimental exploitation so far (60000 plasma seconds). The C flux behaviour in the divertor can be divided in three phases: initial fast drop, moderate reduction phase, and a long lasting phase with almost constant C flux. The Be flux in both divertor legs mirrors the behaviour of C. All experiments and diagnostic techniques demonstrate a strong reduction in C fluxes and C content of more than one order of magnitude with respect to JET-CFC which is in line with the reduction in long-term fuel retention due to co-deposition. There is no evidence of an increase in residual carbon in time, thus no indication that a damage of the thin tungsten coatings on CFC substrate in the divertor occurred.

The JET Active Gas Handling System (AGHS) analytical glovebox contains a custom built Gas Chromatography (GC) system and an Omegatron Mass Spectrometer. The equipment is primarily used to determine...

Detritiation of highly tritium contaminated carbon and metal material used as first wall armour is a key issue for fusion machines like JET and ITER. Re-deposited carbon and hydrogen in the form of flakes and dust can lead to a build-up of the tritium inventory and therefore this material must be removed and processed. The high tritium concentration of the flake and dust material collected from the JET vacuum vessel makes it unsuitable for direct waste disposal without detritiation. A dedicated facility to process the tritiated carbon flake material and recover the tritium has been designed and built. In several test runs active material was successfully processed and de-tritiated in the new facility. Samples containing only carbon and hydrogen isotopes have been completely oxidized without any residue. Samples containing metallic impurities, e.g. beryllium, require longer processing times, adjusted processing parameters and yield an oxide residue. The detritiation factor was 2x10{sup 4}. In order to simulate in-vessel and ex-vessel detritiation techniques, the detritiation of a carbon flake sample by isotopic exchange in a hydrogen atmosphere was investigated. 2.8% of tritium was recovered by this means. (authors)

In the present design of the Storage and Delivery System of the ITER Tritium Plant deuterium, tritium and their mixtures are stored in hydrogen storage beds with a storage capacity of 100 g. During plasma operation it is required that deuterium-tritium gases with well defined ratios of D/T are supplied by the different hydrogen storage beds. Due to the isotope effects the composition of the hydrogen gas mixture supplied by the getter bed may be different from the one absorbed in the getter and may even change during unloading of the bed depending on the variation of the isotope effect with the actual amount of hydrogen isotopes stored in the bed. At the Forschungszentrum Karlsruhe a 1:1 prototype of ITER hydrogen storage bed with a capacity of100 g tritium and a target supply rate of up to 200 Pam3s-1 was designed and manufactured. The getter bed is currently filled with zirconium-cobalt and is installed in an experimental rig coupled with a micro gas chromatograph in order to perform tests under different operation conditions and to characterize the possible isotope effects. In this work a first data on the isotope effect during loading and unloading of the getter bed with the different hydrogen-deuterium mixtures is presented.

A high resolution thermoelectric vacuum calorimeter for tritium measurements is described in detail. The results and experience gained during the measurement of different tritium samples with the vacuum calorimeter as well as comparisons with measurements from other calorimeters are presented. Special specimens have been cut to fit within the measuring volume of 0.5 litre. With the low detection limit of the vacuum calorimeter, tritium in tiles can now be measured by calorimetry for the first time.

Various analytical techniques exist to determine the compositions of gases handled in the fuel cycle of future fusion machines. Gas chromatography was found to be the most appropriate method. The main disadvantages of conventional gas chromatography were the long retention times for the heavy hydrogen species of >30 min. Recent progress in the development of micro-gas chromatography has reduced these retention times to ≈3 min. The usefulness of micro-gas chromatography for the analysis of hydrogen and impurity gas mixtures in the fuel cycle of future fusion machines is presented and the advantages and drawbacks are discussed.

The plasma exhaust processing plant of fusion reactors require on-line analysis of gas streams to monitor various gas compositions. The resolution of an Omegatron increases with decreasing mass and it can easily discriminate between all the 20+ species of ions between masses 1 and 9, except between T+ and He3+ ions. The Omegatron is a candidate device for gas analysis, it is small, simple to operate and relatively inexpensive. The disadvantages are the upper pressure limit of about 1.0E−6 mbar and low output currents, typically less than 1.0E−11 A, which give slow response times. The Omegatron at the Tritium Laboratory Karlsruhe is well established and has been used successfully for qualitative analyses of tritiated gas mixtures. Quantitative measurements on known gas mixtures up to mass 28 have revealed several anomalies, particularly when helium is present. Work to explain this effect is reported. The gas in the mass spectrometer has to be a representative sample and be introduced in a reasonable timescale. A potential pumping and expansion arrangement is described. The Omegatron is highly suited to computer control as the cyclotron frequencies of the various ions are precisely known and techniques to control the Omegatron and evaluate the results are proposed.

At the Tritium Laboratory Karlsruhe (TLK) laser Raman spectroscopy, gas chromatography, mass spectroscopy, calorimetry and ionisation chambers are used to determine the composition of tritium gas mixtures. For the first time a laser Raman experiment was assembled with an actively controlled resonator which yields a 50 times higher Raman signal and with all components (laser, optics, Raman cell and spectrometer) installed inside a glove box. Three gas chromatographs, each with up to six detectors, can determine the gases and their tritiated fractions expected in fusion devices down to the sub-ppm range. Tritium in solids, liquids and gases is determined by means of three calorimeters with a dynamic ranges of up to five orders of magnitude and a lower detection limit of 1 GBq. Since any of these techniques has its shortcomings the best analytical approach is to analyse a sample by more than one method.

The kinetics of the radiation-induced reactions of carbon monoxide with either an equimolar mixture of deuterium and tritium or pure tritium was investigated at several total pressures and within a range of concentrations that simulate those expected in the fuel cycle of a fusion machine, i.e. 1–6% CO in DT or T2 containing 0–8% He. The reaction rate was found to be directly proportional to the initial tritium concentration in the gas mixture. From the initial rate of CO pressure decrease, an average CO consumption yield per ion pair of 1.6±0.8 was calculated from a total of 24 experiments. At a constant composition of CO, DT and He, the rate of CO consumption decreased strongly with decreasing total pressure. The main gaseous products found were per-tritiated methane, per-tritiated higher hydrocarbons, and carbon dioxide. Water and possibly formaldehyde were identified as minor products. Heated palladium/silver from permeators such as used in fusion machine fuel clean-up considerably enhance the radiation-induced reaction rate between tritium and carbon monoxide.

Analytical results of a complete JET cryopump regeneration, including the nitrogen panel, following the first ITER-Like Wall campaign are presented along with the in-situ analyses of residual gas. H/D mixtures and impurities such as nitrogen and neon were injected during plasma operation in the vessel to study radiation cooling in the scrape-off-layer and divertor region. The global gas inventory over the campaign is incomplete, suggesting residual volatile impurities are remaining on the cryogenic panel. This paper presents results on a) residual deuterium on the panel which is related to the campaign very low, b) impurities like nitrogen which sticks on the panel and c) the ammonia production which can be observed in the RGA spectrum., 7 pages, 3 figures. This is an author-created, un-copyedited version of an article accepted for publication in Physica Scripta. IOP Publishing Ltd and IAEA are not responsible for any errors or omissions in this version of the manuscript or any version derived from it

JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ip = 2.0 MA, Bt = 2.0-2.4 T, δ = 0.2-0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges (Paux = 0.5 MW), type III (Paux = 5.0 MW) and type-I ELMy H-mode plasmas (Paux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 × 1021, 1.37 × 1021 and 1.97 × 1021 D s-1 down to 4.8 × 1019, 7.2 × 1019 and 16 × 1019 D s-1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 × 1020 D s-1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device. © 2013 IAEA, Vienna.

Preparatory experiments for the ITER-Like Wall in JET were carried out to simulate the massive Be first wall by a thin Be layer, induced by evaporation of about 2.0 g Be, and to study its impact on fuel retention and divertor radiation with reduced C content and N seeding. Residual gas analysis reveals a reduction of hydrocarbons by one order of magnitude and of O by a factor of 5 in the partial pressure owing to the evaporation. The evolution of wall conditions, impurity fluxes and divertor radiation have been studied in ELMy H-mode plasmas (B t = 2.7 T, I p = 2.5 MA, P aux = 16 MW) whereas a non-seeded reference discharge was executed prior to the evaporation. The in situ measured Be flux at the midplane increased by about a factor of 40 whereas the C flux decreased by ∼50% in the limiter phase of the first discharge with respect to the reference, but erosion of the Be layer and partial coverage with C takes place quickly. To make best use of the protective Be layer, only the first four discharges were employed for a gas balance analysis providing a D retention rate of 1.94 × 1021 D s−1 which is comparable to rates with C walls. But the Be evaporation provides a non-saturated surface with respect to D and short term retention is not negligible in the balance; the measured retention is overestimated with respect to steady-state conditions like that of the ILW. Moreover, C was only moderately reduced and co-deposition of fuel with eroded Be and C occurs. The lower C content leads to a minor reduction in divertor radiation as the reference phase prior to seeding indicates. N adds to the radiation of D and the remaining C, and the N content rises due to the legacy effect which has been quantified by gas balance to be 30% of the injected N. C radiation increases with exposure time, and both contributors cause an increase in the radiated fraction in the divertor from 50% to 70%. The radiation pattern suggests that N dominates the increase in the first discharges though C is still the dominating radiator. Therefore, the validity of a proxy of the Be first wall by a thin Be layer is limited and restricted to plasma operation directly after the Be evaporation.

Migration of 13C has been investigated at JET by injecting 13C-labelled methane at the outer divertor base at the end of the 2009 campaign. The 13C deposition profiles on carbon fibre composite divertor tiles were measured by secondary ion mass spectrometry and Rutherford backscattering techniques. 13C was mainly deposited near the puffing location on the outer divertor base tiles. High amounts of 13C were also found at the outer vertical target: at the bottom of the lower and at the top of the upper plates. Thirty-three percent of puffed 13CH4 was instantly pumped out by the divertor cryopump, which is close to the pump duct entrance. Global 13C transport in the torus was modelled by the EDGE2D/EIRENE and DIVIMP codes, and local 13C migration in the vicinity of the injection location by the ERO code. The DIVIMP and EDGE2D simulations show strong prompt deposition of 13C directly adjacent to the injection point as well as in the far scrape-off layer (SOL) along both the inner and outer divertor targets. In addition, the measured 13C deposition along the outer divertor wall tiles is qualitatively reproduced. However, EDGE2D and DIVIMP do not predict any deposition along the divertor surfaces facing the private plasma on the inner floor tile and inboard of the outer strike point on tile 5. The ERO calculations also indicate that most of the deposition occurs close to the injection location on the vertical face of the LBSRP tile and the horizontal part of tile 6.