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During 1997 JET operation with D-T plasmas, 35 g of tritium were introduced into the torus, mainly by gas puffing. It was found that during this period, the torus tritium inventory would accumulate at a rate of about 40% of the input. After tritium operation ceased, the experimental program continued with deuterium- and hydrogen-fuelled experiments, during which time the tritium inventory decreased to about 17% of the total input. Techniques aimed at detritiation of the torus included methods using deuterium gas (such as deuterium pulsing) which were used in the middle of the experimental campaign, and methods which could adversely affect the torus vacuum conditions (such as air purges) which were reserved for the period after the experimental campaign. Whilst it was found that the plasma tritium fraction could be reduced to below the 1% level in a few days, the tritium inventory reached a virtually steady level of about 6 g by the end of the campaign.

During 1997 JET was operated for an extensive period using a D–T mixture (DTE1). Changes in the design and operation of diagnostic systems made over the years in preparation for this phase are described. A number of diagnostic techniques have been deployed to measure the deuterium and tritium content of the plasma during DTE1 and their results are compared. All diagnostics with a direct vacuum interface with the main vessel have been fitted with tritium compatible pumps and their exhaust gases have been re-routed to the active gas handling plant. All items on the torus which could lead to a significant leak in the event of failure, were required to have double containment. Therefore, all windows, and a majority of bellows and feedthroughs, were designed and installed with a double barrier. Heated fibre hoses were installed to transmit plasma light beyond the biological shield for spectroscopic purposes. Blind fibres and fibre loops were also installed to study the effects of higher neutron fluxes on these fibres. A radiation-hardened video camera was installed to monitor the plasma during the DTE1 discharges. Extra shielding was installed on a number of diagnostics to deal with the higher neutron fluxes during DTE1. The effect of neutron radiation on electronics in the Torus Hall was studied. During DTE1 the tritium fraction was measured at the edge and in the core using several diagnostic methods. High resolution Balmer α line spectroscopy gave a measurement typical of the plasma edge region. In the JET sub-divertor volume the tritium concentration of the neutral gas was measured using Balmer α spectroscopy of a Penning gauge discharge. Using Neutral Particle Analysis, the tritium concentration was measured typically in a zone 20–40 cm from the plasma edge. Local core measurements of the tritium fraction have been made using active Balmer α charge exchange spectroscopy. The error on this measurement is, however, large,∼30%. After the discharge the tritium fraction of the exhaust was measured using the exhaust monitoring system. Using short deuterium neutral injection pulses allowed neutron rate measurements of the tritium concentration in the core region. A further technique used the measured neutron rate and calculated neutron rate from other plasma parameters to determine the tritium concentration.

The neutral tritium concentration in the subdivertor region of JET [P.-H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion 25, 1011 (1985)] is measured during deuterium-to-tritium changeover experiments with a novel species-selective Penning gauge coupled to a high-resolution spectrometer. The subdivertor measurements, when compared with edge and strike point values, are a sensitive characterization of the status of the wall saturation. The neutral transport code (EIRENE) [D. Reiter, Forschungszentrum, Juelich: Report Juel-2599 (1992)] and a wall hydrogen trapping and diffusion code (WDIFFUSE) [J. Hogan, R. Maingi, P. Mioduszewski et al., J. Nucl. Mater. 241, 612 (1997)] evaluate the wall tritium recycling coefficients (RT) and compare them with quantitative, testable models for the dynamic exchange between recycling surfaces and the edge/pedestal region. Using the dynamic inventory model, tritium transport following the injection of trace amounts of tritium has been investigated for JET high-confinem...

Trace amounts of tritium gas have been injected in short puAs into JET ELMy H-modes with a wide range of deuterium gas-fuelling rates. Analysis of the subsequent time evolution of the neutron profile and extraction of the particle transport coeAcients have allowed us to distinguish between broad classes of mechanism which have been suggested as explanations for the H-mode density limit. The high penetration probability (20%) and rapid transport (sE) of fuel ions are shown to be only weakly influenced by strong gas fuelling ‐ hence mixture control is possible even when the total electron content is clamped. ” 1999 JET Joint Undertaking, published by Elsevier Science B.V. All rights reserved.

JET has completed a series of experiments in the Mk I and Mk IIA divertors on the effects of increased geometrical closure and target orientation. The potential benefits from closure were expected to be enhanced volumetric energy loss in the divertor (detachment), increased divertor neutral pressure for better pumping and He exhaust, and reduced main chamber neutral pressure for reduced sputtering. The expected effects on neutral pressures were observed. In ohmic and L-modes this led to detachment at lower upstream density and reduced density limits, in qualitative agreement with code calculations. The pumping speed was increased by about a factor of three. Zeff did not reduce, despite the reduced main chamber neutral pressure. In ELMy H-modes the effects of closure were less distinct, which may have been due in part to ELMs striking the upper surfaces of the divertor and main chamber limiting surfaces. The density limit and confinement quality were unaffected by changes in divertor geometry. Increasing triangularity increased the density limit, but also raised Zeff. Confinement was degraded by either deuterium puffing or nitrogen puffing. Detachment occurred at the inner target between ELMs, but not at the outer target until confinement was strongly degraded. Vertical target ELMy H-modes have thinner SOL's and lower midplane separatrix densities than those run on horizontal targets in Mk IIA. Given the JET observations on the lack of sensitivity of core plasma ELMy H-mode performance to divertor geometry, it appears appropriate to review the possibility of simpler, lower cost divertor options than the deep divertor design currently proposed for ITER.

Experiments to study the exhaust of noble gases (helium and neon) with cryopumping in DIII-D (Advanced Divertor Program (ADP) configuration) and in JET (Mk1 configuration) found significant differences in the global exhaust rate of helium, while efficient neon exhaust was observed in both machines. An attempt to better understand the basic processes governing the exhaust of noble gases in ELMy H-mode with cryopumping has been undertaken. Since divertor geometries investigated in the DIII-D and in the JET cases have significant differences, a comparative modeling study has been undertaken using the MIST core impurity transport code and the b2.5 time-independent divertor transport code. Photodiode measurements are used to determine ELM frequency, and charge-exchange recombination (CER) measurements are compared with the MIST ELM model to evaluate transport coefficients in the core plasma. A significant reduction in the anomalous diffusivities is found for the non-ELM component of radial transport without the need for a pinch velocity, and the model provides a more coherent description than the conventional ELM-averaged approach. Sensitivity to boundary conditions has been studied through establishment of a database of divertor enrichment cases using b2.5.

JET's 1997 Deuterium Tritium Experiment (DTE1) allows a detailed study of hydrogenic isotope recycling and retention in a pumped divertor configuration relevant to ITER. There appear to be two distinct forms of retained tritium. (1) A dynamic inventory which controls the fueling behaviour of a single discharge, and in particular determines the isotopic composition. This is shown to be consistent with neutral particle implantation over the whole vessel surface area. (2) A continually growing inventory, which plays a small role in the particle balance of a single discharge, but ultimately dominates the hydrogenic inventory for an experimental campaign comprising thousands of pulses. This will be the dominant retention mechanism in long-pulse devices like ITER. The JET retention scaled-up to ITER proportions suggests that ITER may reach its tritium inventory limit in less than 100 pulses.

On the basis of experiments carried out in JET in D:T mixtures varying from 100:0 to 5:95 and those carried out in hydrogen plasmas, the isotopic mass dependence of ELM parameters and the edge pedestal pressure in NBI and ICRF heated H mode plasmas is presented. The ELM frequency is found to decrease with the atomic mass number in both ICRH and NBI discharges. However, the frequency in the case of ICRH is about 8-10 times higher than that in the case of NBI. Assuming that ELMs occur at a critical edge pressure gradient, limited by the ballooning instability, the scaling of the maximum edge pressure is most consistent with the assumption that the width of the transport barrier scales as the ion poloidal Larmor radius governed by the average energy of fast ions at the edge. The critical edge pressure in NBI heated discharges increases with the isotopic mass, which is consistent with the higher deduced width of the edge transport barrier in tritium than those in deuterium and hydrogen. The critical edge pressure in ICRH discharges is smaller, presumably, due to the smaller fast ion contribution to the edge region. As a consequence of the edge pressure scaling with isotopic mass, the edge operational space in an ne-Te diagram increases with operation in tritium. If the evidence that the edge pedestal width is governed by the average energy of fast ions in the edge prevails, the pedestal in ITER would be controlled by the slowing down energy spectrum of alpha particles in the edge.

An overview of JET experimental results in DT plasmas directly relevant to ITER modes of operation is presented. Experiments in D:T mixtures varying from 100:0 to 10:90 and those carried out in hydrogen plasmas show that the H mode threshold power has an approximately inverse isotope mass dependence. Matching some of the key dimensionless parameters to the ITER values, the ITER similarity experiments with ITER shape and safety factor q show that the global energy confinement time is practically independent of isotopic mass (~A0.03±0.08), where A is the atomic mass of the hydrogenic species. Subtracting the edge pedestal energy (which scales as ~A0.57±0.2) from the total stored energy leads to a ~A-0.17±0.1 dependence of confinement in the plasma core, very similar to that expected from the gyro-Bohm transport (~ A-0.2) model. The observed scaling of the edge pedestal energy is consistent with a model in which the edge pressure gradient saturates at the ballooning limit over a region of width that scales as the ion poloidal Larmor radius governed by the average energy of the fast ions in the edge. The steady state total stored energy for a given input power in both ICRH and NBI discharges is the same despite the lower edge pedestal in the ICRH case, which is compensated for by more peaked power deposition profiles in ICRH. The ELM frequency is smaller with NBI; it decreases with isotopic mass in both NBI and ICRH discharges. A steady state, type I ELMy H mode discharge with ITER shape and q at 3.8 T/3.8 MA with an input power of 22 MW produced a Q ≈ 0.18 for 3.5 s and extrapolates well to ignition with ITER parameters. Here, Q is the ratio of fusion output power to input power. The thermal ELMy H mode confinement in both deuterium and tritium gas fuelled plasmas decreases significantly when the plasma density exceeds 0.75 of the Greenwald (nGW) limit, and the maximum density achieved is 0.85nGW. In L mode, the density limit decreases with increasing isotope mass roughly in accordance with code predictions. ITER reference ICRH scenarios have been evaluated. Second harmonic heating of tritium at the densities available in JET produces strong tails and heats electrons predominantly as expected. The 3He minority in 50:50 D:T and tritium dominated plasmas showed strong bulk ion heating leading to ion temperatures up to 13 keV with ICRH alone. Deuterium minority ion cyclotron heating in tritium plasmas at a power level of 6 MW produced steady state record values of Q ≈ 0.22 for more than 2.5 s.

Results of experiments investigating the performance of the JET Mark IIA divertor are reported and compared with the performance of its Mark I predecessor. The principal effect of reducing the divertor width (increasing closure) was to increase pumping for both deuterium and impurities while reducing upstream neutral pressure. Neither the orientation of the divertor target relative to the divertor plasma nor the width of the divertor had a major influence on core plasma performance in ELMy H modes. Changing the core triangularity and thus the edge magnetic shear modifies the ELM frequency in ELMy H mode plasmas, thereby changing the peak divertor power loading. The integrated performance of the core and divertor plasmas is reviewed with a view to extrapolation to the requirements of ITER. The confinement of JET ELMy H modes with hot, medium density edges is good (H97 ≈ 1) and follows a gyro-Bohm scaling. The impurity content of these discharges is low and within the ITER requirements. When an attempt is made to raise the density with deuterium gas fuelling, the ELM frequency increases and the confinement, especially in the edge, decreases. Good confinement can be achieved in JET either by producing a large edge pedestal, typically in discharges with NB heating or by centrally peaked heating with ICRH schemes. Large amplitude type I ELMs, which are present in all discharges with a large edge pedestal, would result in unacceptable divertor plate erosion when scaled to ITER. Since the power deposition profile due to α heating in ITER is calculated to be intermediate between the JET NB and RF heating profiles, it is likely that operation in ITER with small ELMs in order to reduce first wall loading will result in degraded confinement compared with present day scaling laws.

The experimental characteristics of divertor detachment in the JET tokamak with the Mark?I pumped divertor are presented for ohmic, L?mode and ELMy H?mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes, as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment, will be described. The observed detachment characteristics, such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer (SOL), and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and D? emission from the divertor region, will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree of detachment and the window of detachment is proposed, and all the observations of the JET Mark?I divertor experiments are summarized in the light of this new quantitative definition of divertor detachment.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

First measurements are presented of the total power loss through neutral particles and their average energy in the JET divertor. The method used distinguishes between the heat flux and the electromagnetic radiation on bolometers. This is done by comparing measurements from inside the divertor either with opposite lines of sight or with a tomographic reconstruction of the radiation. The typical value of the total power loss in the divertor through neutrals is about 1 MW. The average energy of the neutral particles at the inner divertor leg is 1.5–3 eV when detachment is in progress, which agrees with EDGE2D/NIMBUS modelling.

Profiles of Te and Jsat measured by Langmuir probes at the target plates of the JET MkI divertor and the target plate power loading are modelled via the EDGE2D/NIMBUS codes. The low-density ( n e ∼ 1 × 10 19 m −3 ) OH plasma preceding high-power NBI, as well as the ELM-free high performance H∗ and high performance rollover HRO phases are considered. Experimentally, the power splitting between ions (Pi ≥ 7 MW) and electrons (Pe n e can increase up to a factor of three as the H∗-phase progresses. In modelling χi is taken as 1 m2/s. To match the divertor Te-profiles, χe must be varied from the strike point outwards as ∼ 0.5 → 3 m2/s in all phases. Perpendicular particle transport is assumed to be purely diffusive or with a pinch. Duplication of Jsat(R) requires: D ⊥ ∼ 0.03 → 0.015 m 2 /s (OH → H ∗ ) , or ν pinch D ⊥ ∼ 15 → 45 → 25 m−1 (OH → H∗ → HRO) with D⊥ = 0.1 m2/s. Sensitivity studies involving different recycling scenarios, deep and shallow computational grids, wall material and the thermal transport barrier are carried out to judge the possible influences on the deduced transport coefficients.

Detailed experiments have been carried out with the JET Mark 1 pumped divertor to characterise high recycling and detached plasma regimes. This paper presents new measurements of high resolution divertor ion flux profiles that identify the growth of additional peaks during high recycling discharges. These ion flux profiles are used in conjunction with Dα and neutral flux measurements to examine the physics of divertor detachment and compare against simple analytic models. Finally, problems are highlighted with conventional methods of single and triple probe interpretation under high recycling conditions. By assuming that the single probe behaves as an asymmetric double probe the whole characteristic may be fitted and significantly lower electron temperatures may be derived when the electron to ion saturation current ratio is reduced. The results from the asymmetric double probe fit are shown to be consistent with independent diagnostic measurements.

Retention of neutral deuterium and nitrogen in the JET MK I divertor has been investigated. Results show that ohmic plasma detachment reduces deuterium retention, that the magnetic divertor configuration has some influence on the achievable deuterium retention, and that nitrogen in nitrogen-seeded steady state detached H-mode discharges accumulates in the divertor.

An extensive experimental investigation of reactor-relevant features of H-mode plasmas has been carried out using the JET-pumped divertor. Steady-state H-modes have been established with high confinement and with durations of up to 20 s. At the highest densities, these plasmas return to the L-mode when the radiated power fraction is %. Detached H-modes have been produced by gas-puffing deuterium plus nitrogen, and their properties studied at total input powers of up to 32 MW. Helium transport and exhaust have also been investigated, and a systematic comparison of H-modes using carbon fibre composite and beryllium divertor targets has been carried out.

The results of a series of ITER simulation experiments on JET are described. A series of H-mode threshold experiments are shown to reproduce one of the standard power threshold scaling expressions that is being used to predict the power threshold in ITER. Then, from a series of experiments in which the Larmor radius scaling of ELMy H-modes is examined, it is concluded that the scaling of the confinement is gyro-Bohm-like provided the power levels are well above the threshold. Finally, we show that at high there is a dramatic reduction in the confinement at .

A model based on a local mixing model has been previously developed at JET to explain the recovery of tritium after the first PTE experiment. This model is extended by a 0D plasma particle balance model and is applied to data from Tore Supra wall saturation experiments. With only two free parameters, representing the diffusion of hydrogen atoms and the volume recombination process between hydrogen atoms into molecules, the model can reproduce experimental data. The time evolution of the after-shot outgassing and the integral amount of particles recovered after the shot (assuming 13 m 2 of interacting surfaces between plasma and walls) are in good agreement with the experimental observations. The same set of parameters allows the model to simulate after-shot outgassing of five consecutive discharges. However, the model fails to predict the observed saturation of the walls by the plasma. Results from helium glow discharge (HeGD) can only be partially described. Good agreement with the experimental hydrogen release and its time evolution during HeGD is observed, but the model fails to describe the stability of a saturated graphite wall.

During the JET Preliminary Tritium Experiment (PTE), an estimated 2 × 1012 (1.1 × 1021 atoms) of tritium were injected into the JET vacuum vessel. A series of experiments was performed whose purpose was to deplete the torus of tritium, to compare the effectiveness of different methods of tritium removal and to obtain a quantitative understanding of the processes involved. The effectiveness of the cleaning procedures was such that the normal tokamak programme was resumed one week after the PTE and the routing of exhaust gases to the atmosphere after two weeks. The release of tritium from the vessel was found to scale with the deuterium release from the vessel, suggesting that dilution and mixing of the hydrogen isotopes in the vessel walls is important. High density, disruptive tokamak discharges were found to be the most successful plasma pulses for tritium removal. Purges with deuterium gas were also effective and have the advantage of operational simplicity. Helium discharges, on the other hand, resulted in low tritium release from the vessel walls. It was demonstrated that the tritium release rate could be predicted using data from hydrogen to deuterium changeover experiments. Using the superior quality of data available from the tritium cleanup experiment, the physical mechanisms necessary to describe the hydrogenic uptake and release from the JET torus were identified. Tile release of tritium is reproduced using a model that incorporates implantation into a thin surface layer as well as diffusion of tritium into and out of the bulk material

The dependence of impurity production and retention on the divertor density, on the power flow into this region as well as on the X-point to target distance are investigated. Model predictions suggest a good impurity retention above a certain divertor (scrape-off) density threshold, which is dependent on heating power. In our experiments pre-programmed midplane or X-point gas puffs were used to scan the density, as well as to avoid the depletion of particles from the divertor and the scrape-off during H-modes. The gas puffs reduce T e and increase N e in particular at the outer strike zone. In general the Be as well as the C influx increases with density, which is understood from the T e ( T i ) dependence of the sputtering yields. The impurity retention shows the expected improvement with increasing scrape-off (divertor) density as well as with increasing X-point to target distance (connection length).

The first tritium fuelled discharges in JET were followed by a series of cleanup discharges aimed at removing tritium from the torus. Measurements of tritium in the plasma and the exhaust were used to assess the efficiency of the various cleanup techniques. A preliminary estimate of the in-vessel tritium inventory is presented.

A study of density control in JET has revealed that dynamic retention in the beryllium surface of the first wall is an important factor in controlling recycling. A self-consistent global particle balance model that simulates the recycling in JET has been devised. The assumption is that particle transport is effectively diffusive in the materials encompassing the plasma and particle release is by recombination to molecules. The effect of neutral particle screening in the plasma edge is also globally taken into account. Comparison with experimental results of discharges of normal duration (20 s) shows good agreement. Discrepancies appear for 60 s long discharges, suggesting changes of material properties.

Isotope exchange experiments have been performed in JET using hydrogen, deuterium, and, in the recent preliminary tritium experiment (PTE), tritium. The rate of change-over from one isotope to another involves two quite different time constants. We have modelled this behaviour using a multireservoir model which splits the accessible hydrogenic particles into two groups, each having a different rate of exchange of particles with the plasma. By applying this model to the sequence of discharges during and after the PTE, we can determine the parameters in the model. The resulting fit also gives a good representation of hydrogen/deuterium change-over experiments, indicating that the tritium behaves in the same manner as other hydrogen isotopes, at least as far as recycling is concerned. Discrepancies between the model and the actual measurements of tritium recovery after the PTE lead us to conclude that isotope exchange processes resulting from collisions of molecules with the vessel walls play a significant role in spreading tritium around the machine.

Measurements have been made of the tritium recovery from the JET pumping system both after high power neutral beam injection where the tritium is produced from D(d, p)T reactions and after tritium is injected into the torus in the first tritium experiment (FTE). The tritium has been detected using a liquid scintillation counter which has allowed measurements to be made over a wide range of concentrations. The recovery rate has been found to follow a power law R=At−n over the period from 10 to 3000 s after each discharge. The value of n varies from 0.8 when tritium is first introduced into the plasma to ∼0.5 many discharges later. This effect is attributed to a change in the depth distribution of the tritium in the surface.

Deuterium recycling during and after ohmic limiter discharges in the Be-phase of JET are compared with that during the C-phase. For the Be-phase, effective transport parameters for diffusion and recombination of deuterium at surfaces of in-vessel components of JET are derived using plasma discharge data only. This and the analysis of the outgassing behaviour after the discharge suggest that deuterium release is controlled by recombination with D/k (D is the diffusion, k the recombination coefficient) of about 1021−1022 m−2, larger than expected from laboratory measurements.

Obtaining precision measurements of the relative concentrations of hydrogen, deuterium, tritium, and helium in the divertor of a tokamak are an important task for nuclear fusion research. Control of the deuterium-tritium isotopic ratio while limiting the helium ash content in a fusion plasma are key factors for optimizing the fuel burn in a fusion reactor, like the International Tokamak Experimental Reactor (ITER). A diagnostic technique has been developed to measure the deuterium-tritium isotopic ratio in the divertor of the Joint European Torus (JET) with a species-selective Penning vacuum gauge. The Penning discharge provides a source of electrons to excite the neutral hydrogen isotopes in the pumping duct. Subsequently, the visible light from the hydrogen isotopes is collected in an optical fiber bundle, transferred away from the tokamak into a low radiation background area, and analyzed in a high resolution Czerny-Turner spectrometer, which is equipped with a fast charge coupled device (CCD) camera for optical detection. The intensity of the observed line emission (D{sub {alpha}} -- 6561.03 {angstrom}; and T{sub {alpha}} -- 6560.44 {angstrom}) is directly proportional to the partial pressure of each gas found in the divertor. The line intensity of each isotope is calibrated as a function of pressure.more » The ratio of the line intensities thus provides a direct measurement of the deuterium-tritium isotopic ratio. The lower limit for the determination of the deuterium-tritium isotopic ratio is about 0.5%. The applicable pressure range for this system is from 10{sup {minus}5} mbar to a few times 10{sup {minus}3} mbar.« less

In order to evaluate the deuterium and hydrogen distributions in JET after long term operation in deuterium in 1985, as well as after a few weeks with deuterium at the end of the hydrogen-based operations in 1984, limiters have been analysed using nuclear reaction techniques and laser induced mass analysis (LIMA). The results show the presence of very high amounts of deuterium and 1H (up to more than 1022 atoms/m2 on the sides and edges of the limiters). If this amount were the result of the implantation of deuterium into the limiter graphite, it would require ion energies far higher than those expected in the boundary layer. The deuterium concentration shows a broad minimum near the centre of the limiter, where the heat load from the plasma is a maximum. The depth distributions, which are obtained by the 3He(d, p)4He nuclear reaction, indicate that deuterium is present to a considerable depth on the sides of the limiters. Similar amounts have been seen at the inner wall of the vessel — either on small samples mounted on the wall, or on the carbon tiles which covered the wall in 1985 — and on the RF antennae. It appears that co-deposition of carbon and deuterium enables significant trapping of hydrogen isotopes on all exposed surfaces.

Deuterium is found all over the interior surfaces of JET, at areal densities of up to 10 22 atoms m −2 . From all the analyses carried out on samples removed after each operational campaign, maps of the distribution of deuterium retained in the vessel and estimates of the total deuterium inventory can be obtained. Results from the operational periods 1985, 1986 and June-August 1987 are compared and show that the amount and pattern of retention has changed markedly. In particular the percentage retention has increased from ~ 4% in 1985/6 to ~ 40% or more in 1987, following installation of belt limiters and greatly increased areas of graphite wall tiles. The retention has implications for future tritium operations, and some of the retained gas is released on venting the vessel.

Trapping of helium after implantation at energies of 8 to 150 keV and fluences up to 1019 He-ions/cm2 in nickel at room temperature is studied by measuring the thermal desorption spectra during linear heating up to 1000°C. At several annealing stages the trapped helium is measured by means of the nuclear reaction 3He(d, α)H and the target surface is observed by laser scattering and with the scanning electron microscope. The thermal desorption spectra depend strongly on the implantation fluence but only slightly on the implantation energy, indicating a similar trapping of He in the lattice for the implantation energies used here, The temperature at which desorption starts decreases with increasing fluence. Above the critical fluence for blistering an additional low temperature (150°C) desorption maximum is found. The desorption peak at 150°C can be approximated theoretically with a single jump desorption process of first order and a Gaussian distribution of activation energies around 1 eV. The mea...

The distribution of metal impurities on the surface of the graphite limiters in JET after exposure to discharges in the operational periods 1983, 1984 and 1985 have been examined. A simple model for erosion and redeposition is proposed which is in broad agreement with the observed distributions. It is concluded that impurity deposition on the limiter from all sources is rapidly eroded and redeposited on the limiter surface during normal tokamak discharges so that the final distribution is dominated by this mechanism. Mechanisms are also identified for net removal of material from the limiter.

Wall long term samples (LTS) which were exposed during the 1984 and 1985 operational periods of JET have been investigated by means of surface analysis techniques. The results indicate that deposition of wall material (Fe, Cr, Ni) and its distribution on the vessel wall were strongly affected by wall erosion on bellows protection plates during unstable discharge phases. Erosion sources and deposition sinks are spatially correlated indicating that most of the eroded material is deposited close to its source.

The determination of actual depth profiles of metals deposited simultaneously with a variety of lighter elements on a rough technical carbon surface is complicated by the unknown deposit topography and possibly by diffusion to substantial depths through pores or into the bulk. It is demonstrated that a suitable combination of RBS and PIXE allows a useful estimate of “effective depths” in the sense in which it may be important for the better understanding of plasma limiter interactions.

The surface layer analysis of collector probes after exposure in plasma experiments may often be obscured by a variety of effects, many of which are usually underestimated or ignored. Examples obtained by RBS and PIXE analysis of various types of carbon probes are used to illustrate some of the problems. However, the investigation and discussion should also have consequences for other analysis techniques and materials. In particular, it is questionable whether the fluxes of chemically active elements, notably oxygen, are measurable with collector probes at all.

8 keV 3 He ions have been implanted into edge-oriented pyrolytic graphite at temperatures of 120 K and 295 K. The retention behavior observed by nuclear reaction and residual gas analyses differs markedly from that observed in metals. Despite the negligible reflection coefficient, the mean re-emission coefficient at low fluence amounts to ~30% at room temperature, with a relative maximum at a fluence of 1.5 × 10 17 cm −2 . Saturation is observed at fluences above 5 × 10 17 cm 12 . The implanted He is completely removed by annealing to 750 K. Reimplantation at room temperature shows an enhanced retention (~90%) at low fluences. The experimental results are discussed in terms of He diffusion, trapping at damage sites and possible structural changes of the material.

Simultaneous measurements have been made of blister formation and gas re-emission during bombardment of Ni with 8–40 keV 3He+ at room temperature. It is shown that re-emission and blister formation start at the same critical fluence. An initial peak in the re-emission rate is correlated to a step in the total number of blisters. A more detailed analysis of the experimental data shows that the area contributing to gas re-emission is much larger than the blister area. This suggests the formation of an interconnected layer below the surface in which cracks at blister sites provide gas emission channels to the surface.

The interaction of the boundary plasma with the limiters has been analyzed for two 5 MA, 20 s Ohmic discharges in JET by Langmuir probes and by an all carbon collector probe. The deposition of D, C, Ni and Cr and the erosion on the probe have been measured as functions of the distance to the last closed magnetic flux surface. The values are compared to the particle fluences and their energies as calculated from the measured plasma parameters. The calculated deuterium fluences are about a factor of 5 to 10 larger than the deposition on the probe and the calculated e-folding length is steeper than the measured values. The carbon deposition and erosion at the probe has been determined with the aid of 13 C implants as well as from the depth of the metal and deuterium depositions and the interference colours of the deposited films. The distribution of deposited carbon is similar to the trapped deuterium with a ratio D : C of about 0.3, indicating that D trapping is dominated by saturation of the deposited C layer. The metal deposition is small; it amounts to a few times 10 −3 of the carbon deposition, indicating the dominance of carbon in the boundary layer.

Gas re-emission and surface structure modifications measured by diffuse scattering of light were observed simultaneously during the implantation of 40 keV helium in poly- and single crystalline nickel at room temperature. At a critical fluence of 5.2 × 1017 He+/cm2 the He concentration reaches a maximum of 0.4 He atoms/Ni atom at the most probable range and gas re-emission and blistering start simultaneously. With implantation up to 2 × 1018 He+ /cm2 the re-emission approaches 100% of the incident flux, the scattered light intensity reaches a maximum and the He concentration reaches about 0.3 He atoms/Ni atom over the whole range. For further implantation the intensity of scattered light decreases probably due to the blister covers becoming removed by sputtering, while the gas re-emission remains unchanged. It can be inferred from the measurements that the blisters are not the dominant source of gas re-emission, but that both effects are due to a common mechanism.

This paper presents a review of hydrogen and helium recycling phenomena in tokamaks with limiters and walls largely made out of carbon (graphite, a-C:H layers). The key points of interest are the plasma fuelling efficiency, the wall pumping phenomena as observed in JET, TFTR, TEXTOR and other machines under various fuelling schemes (gas, neutral beams, pellets), the release of hydrogen/helium from material surfaces during and after plasma discharges and the long term retention (total particle inventory) of hydrogen in graphite or carbonised structures in tokamaks. The effect of a combined hydrogen/helium plasma on recycling is also discussed. It is shown that only part of the above phenomena can be understood in terms of processes between hydrogen/helium and carbon as known from simulation experiments (ion beams, gas discharge facilities) and that others (in particular the JET wall pumping phenomenon) have still to be explained. Possible mechanisms are outlined and discussed by means of global models.

Isotope exchange experiments from hydrogen to deuterium in JET are theoretically described by employing a simple global isotope exchange model. Experimental results for discharges with limiter temperature around 250 °C can be approximated by this model if an additional slow diffusion process of hydrogen in the limiter bulk is assumed. In discharges where thermal desorption occurs due to higher limiter temperatures (≳1000°C) (post carbonisation discharges) the change over process seems to be predominantly governed by thermal processes.

A Two-Reservoir Model for recycling of hydrogenic atoms at graphite surfaces in JET is presented. Analysis of density transients induced by Neutral Beam Injection (NBI) and other modes of operation are used to determine the parameters of the model. In addition to describing many observed features of transient density and recycling behaviour, the model also reproduces an asymptotic external fuelling efficiency, ϵ s ≈ 1 s, for H/D flux incident on graphite, is determined. Stronger, saturable pumping observed after helium-discharge conditioning is also described. The importance of these pumping mechanisms for control of density and its profile, and for optimization of Q dt in JET, is discussed.

This paper describes a 1.6 MV tandem accelerator-based facility which has been used to measure deposit levels on collector probes that are exposed to plasma discharges in the JET tokamak. The region probed is at the edge of the plasma, of which the central temperature and current may be ∽108 K and 7 MA, respectively. The probes are transported entirely under ultrahigh vacuum conditions to the analysis chamber, wherein Rutherford backscattering, Nuclear reaction analysis, elastic recoil detection and particle induced X-ray emission may be performed. A second chamber in the same vacuum system offers the more surface specific techniques of Auger electron spectroscopy and X-ray photoelectron spectroscopy. The complete system enables the determination of time-resolved information (during the plasma pulse) and radial profiles of all the elements of interest (such as hydrogen, deuterium, carbon, oxygen, chromium, iron and nickel). The particular example cited is a study of erosion/deposition during a 3 MA ohmically heated reference discharge.

The amounts of H and D trapped in the surface layers of samples made out of C, Ni, Inconel and Si, installed at the vessel walls of JET and removed after the 1984 and 1985 periods of operation were measured. After the 1984 operation period the D concentration in carbon probes was of the order of 5 × 10 19 /m 2 . while the hydrogen was about a factor of 40 higher. This can be attributed to a final glow discharge cleaning in hydrogen at the end of the operation period. After the 1984 period of operation the amounts of deuterium and hydrogen were nearly equal at about 5 × 10 21 /m 2 . A comparison of the amounts of deuterium and hydrogen with the amount of carbon deposition on the Inconel wall samples shows a ratio of about 0.3 to 0.4 (H + D) per C-atom. The hydrogen concentrations trapped in the vessel wall correspond to more than 100 times the amount of hydrogen isotopes in the plasma during a discharge.

The surface layers of tiles from the carbon limiters in octant 1, ocant 4 and the RF-antenna limiters were analysed after the end of the first three experimental sessions of JET, 1983, 1984 and 1985. On all tiles metallic contaminants were found from the vessel walls and the RF antennae. In general on all areas deposition and erosion was observed. At the part of the tiles in direct contact with the plasma the erosion dominates, whereas deposition dominates on the outer parts of the limiter radially further away from the plasma. In the small gaps between adjacent limiter tiles, deposited layers of carbon having a thickness of several μm were found, which were partly flaked and contained metallic impurities. Electrical are tracks were observed on the limiter tiles and in the area between limiter tiles.

The modifications of the vessel wall and limiter surfaces in today's large plasma machines are reviewed with special emphasis on the investigations at JET concerning material erosion, transport and redeposition. This is correlated with the problems of impurity release and hydrogen recycling. The specific roles of the limiters and the vessel walls are discussed and some consequences on the use and the distribution of different materials for in-vessel components in tokamaks are indicated.

The JET scrape-off plasma during X-point discharges with L- and H-modes is described. There is a flow of current in the scrape-off layer which can influence equilibrium, transport, heating and impurity generation in the scrape-off layer. A local measurement of D ⊥ is obtained in the divertor region. The saturated flux of ions flowing parallel to the magnetic field, I SAT is ∼10 A/cm2 at the divertor separatrix in ohmic discharges. In L-mode with NBI it increases by a factor ⩽ 2, depending on injected power level ( P NBI MW ) and decreases with H-mode. The e-folding decay length, λ I SAT , is the same in ohmic and L-mode but decreases with H-mode. The electron temperatures at the divertor target are ~ 30 eV (OH) and ~ 50–60 eV (H-mode). Measurements of T e during L-mode are not reported because of the high level of fluctuations. An enhancement in the erosion of carbon due to ion-stimulated desorption at the divertor target is discussed. Flow reversal in the scrape-off layer is demonstrated.

The measurement of the relative concentrations of hydrogen, deuterium, tritium and helium is an important task in the nuclear fusion research area. Control of the deuterium-tritium (D-T) isotropic ratio and limiting the helium ash content in a fusion plasma are the key to optimizing the fuel burn in a fusion reactor such as ITER. A diagnostic technique has been developed to measure the D-T isotopic ratio in the divertor of a tokamak with a Penning vacuum gauge. The Penning discharge provides a source of electrons to excite the neutral deuterium and tritium in the pumping duct. Subsequently, the visible light from the hydrogen isotopes is collected in an optical fibre bundle, transferred away from the tokamak into a low radiation background area and detected in a high resolution Czerny-Turner spectrometer, equipped with a fast CCD (charge-coupled device) camera for optical detection. The intensity of the observed line emission (D α , 6561.03 A; T α , 6560.44 A) is directly proportional to the partial pressure of each gas found in the divertor. The line intensity of each isotope is calibrated as a function of pressure. The ratio of the line intensities thus provides a direct measurement of the D-T isotopic ratio. The lower limit for the determination of the D-T isotopic ratio is about 0.5%. This system is applicable for the pressure range from 10 −5 mbar to a few times 10 −2 mbar.

Obtaining precision measurements of the relative concentrations of hydrogen, deuterium, tritium, and helium in the divertor of a tokamak is an important task for nuclear fusion research. Control of the deuterium–tritium isotopic ratio while limiting the helium ash content in a fusion plasma are key factors for optimizing the fuel burn in a fusion reactor, like the International Tokamak Experimental Reactor. A diagnostic technique has been developed to measure the deuterium–tritium isotopic ratio in the divertor of the Joint European Torus with a species-selective Penning vacuum gauge. The Penning discharge provides a source of electrons to excite the neutral hydrogen isotopes in the pumping duct. Subsequently, the visible light from the hydrogen isotopes is collected in an optical fiber bundle, transferred away from the tokamak into a low radiation background area, and analyzed in a high resolution Czerny–Turner spectrometer, which is equipped with a fast charge coupled device camera for optical detection...