9th ITPA meeting on SOL/divertor physics

The meeting was held over the period May 7-10 at the Max Planck Institut fur Plasmaphysik in Garching Germany. The local coordinator was Arne Kallenbach. The meeting lasted 3-1/2 days and concentrated primarily on gas retention and modelling (see agenda in Appendix A). A number of sessions were held jointly with the Pedestal group. There were over 50 participants (Appendix B).

There is an executive summary (section I) followed by a more summaries of the sessions (section II) and the individual talks (section III).

I. Executive Summary The primary foci of this meeting were discussions and presentations on the ITER program, first-wall and performance specifications, D/T retention, and the state of SOL/divertor modelling. During the ITER discussions our main thoughts and concerns regarding ITER, first discussed during the 8th SOL/divertor meeting, were reiterated: It is very important that the ITER first wall be redesigned to include easily replaceable protection limiters so that a) the heat loads will be in known locations, and b) those locations can be repaired/replaced. There was a consensus that ITER should have the capability for at least one full exchange of the first wall material. While the group could not agree on a particular material, or set of materials for ITER PFCs, it was agreed that at some stage ITER should test a fully DEMO/reactor relevant material combination for both first walls and divertors, likely tungsten. There was general consensus that the current allocation of time for ITER’s hydrogen phase was too short to properly get all diagnostics, control systems, and first-wall PFCs ready for the D/T phase. During that period a strict requirement will be that remote handling will be required to fix any problems. There was also a general consensus that the H-phase is the time to test our physics understanding and engineering hardware; gas retention studies are needed to determine if there if H retention is too high and to develop T removal techniques. Additional power is needed to achieve H-mode confinement and high stored energies; This will allow testing of first-wall and divertor components as well as disruption and ELM mitigation techniques. We devoted a session to comparison of the various D retention measurements which give a wide range of D retention as a fraction of injected gas. The primary techniques are post-campaign Nuclear Reaction Analysis (NRA) of tiles (giving 3-5% retention of injected gas) and discharge-integrated global measurements of gas injected and gas removed from the chamber (giving 20-50% retention of injected gas). While the former is a localized measurement it integrates over long periods of cleaning as well as disruptions – thus leading to a minimum value for retention. The latter technique is more representative of what can happen during a discharge but fails to properly measure what can be removed between discharges. Several methods of extrapolating these measurements to ITER were discussed and the difficulties therein (effect of the more frequent disruptions and shorter pulses in most of current tokamaks, differences in PFC material to ITER…) and are the subject of future comparisons among tokamaks. But at present even if ITER T retention corresponds to the minimum retention fraction the T retention will still be high enough to stop operation for its removal. We hope to better measure the local retention as a function of incident ion flux, a quantity more easily scaled to ITER.

Recent studies have shown that another process in addition to D co-deposition with carbon can lead to significant D retention in current devices, so-called long-range retention; D is found deep within the PFC surface (~ 10 microns, the limit of measurement depth). Such retention will be more difficult to remove than from surfaces as the entire tile must be heated. Presentations of laboratory research and modelling of this process pointed to the same effect for tungsten and Mo – namely ion fluxes into the surface are higher than D can recombine into molecules at the surface. This appears to lead to a buildup of D pressure within the surface and atomic dislocations which then diffuse into the solid providing sites for the D to reside. The analysis of a similar effect for CFC carbon tile indicates that the deep retention there occurs due to C:D molecule migration down pores in the 3D weave of the tiles. Operation of any PFC material in a nuclear environment leads to creation of T trap sites deep within the material proportional to the neutron fluence; the group was very concerned about this possibility. The understanding of these processes for high-Z and carbon PFCs and the effect of neutron damage is clearly important and will continue to be an emphasis of the group. A major emphasis of this meeting was to review the current state of modelling, something we do every several years. A wide range of modelling efforts were presented ranging from interpretive to predictive codes. There are a number of new models being developed including efforts at incorporating our understanding of turbulence into fluid codes, kinetic codes, pedestal fueling, material surface models…. A very important project that was initiated through the ITPA collaborations was that of code-code comparisons. That effort has revealed a host of differences among the major fluid models of the SOL and divertor.. Those differences are primarily in the form of physics assumptions of which the user is usually unaware. During the discussion the consensus of the group was that several key physics questions need to be better modeled as soon as possible - detachment physics, SOL flows (and accompanying electric fields) and kinetic or transient effects. A number of detailed studies of ELM physics were presented. JET ELMs studies indicated that the ELM n number decreased with increasing ELM energy potentially helping to benchmark ELM models. When ELMs impact on the JET inner divertor layers are eroded having implications for a potential source for dust as well as explaining the ability of C:D molecules (or clusters) to cross the separatrix and co-deposit on the dome region. This result may have negative implications for T removal through radiative disruptions. The erosion due to large energy ELMs in JET leads to large effects on the core plasma (increasing radiated power and lowering confinement). ASDEX-Upgrade results clearly show a dependence of ELM power asymmetry on the current flow during the ELM. Several joint sessions were held with the Pedestal group. The main areas of overlap that were covered are ELM mitigation through pellets or ergodization, as well as the role of fueling in determining the characteristics of the pedestal. Pellet-pacing appeared to be of use for ITER although the level of fueling that occurs with pacing needs to be minimized. Of concern for ergodization (besides how to implement it) was the physics question of whether the pedestal density can be sustained at levels corresponding to without ergodization; the density cannot be too low for achievement of Q=10 as well as proper divertor control. It was suggested that present experiments should pursue combined pellet fueling and ergodization. II. Session summaries –

II-1 - Joint session on ELM effects - Chair G. Matthews The problem of ELM effects in ITER is pretty much divided into 4 parts:

1. What leaves the pedestal (energy, particles), 2. What arrives at the various material surfaces of the tokamak (energy density, particle

fluxes and temperatures), 3. How the surface responds (heating, sputtering, sublimation etc.), 4. How does the plasma respond to the impurity influxes (radiation, subsequent ELM

behaviour etc.). This joint session covered mainly items 2-4 of this list. The progress in these different areas from the talks and discussion is summarised below: 2. ELM deposition at surfaces – recent JET studies of ELM n number found it to decrease weakly with ΔWELM. This tendency predicts n >> 1 for small ELMs (eg.Type-III), and n → 1 for giant ELMs consistent with theoretical expectation. If small ELM regimes are utilized in ITER the heat loads would be more spread out? ASDEX studies showed a correlation between ELM power asymmetry and the current flow during the ELM. This asymmetry puts more power on the inner divertor which is favourable for ITER since the steady state power is lower there. However, the correlation with SOL current was not really understood in terms of current theoretical expectation. A JT60-U study was presented correlating Mach flows and electron temperature during and after ELMs. 3. Response of surfaces to ELMs – Using Quartz Crystal Microbalances (QCMs) JET ELMs were shown to mobilise surface layers leading to line of sight migration to remote areas. There was considerable discussion as to whether this eroded flux consisted of particles or molecules or clusters which is of considerable importance for the issue of dust production. The JT60-U results showed evidence for increased chemical sputtering during ELMs. On a positive note high energy ELMs in ITER will tend to reduce the buildup of layers at the inner divertor surface. The effect on buildup of layers in shielded regions (e.g. private flux region) will depend on whether the eroded flux (molecules or clusters) coming from the inner divertor can make it across the separatrix to the dome. 4. Plasma response after the ELM – JET results show a marked increase in divertor radiation for ELMs above 600KJ. It is uncertain whether the radiation is due to ablation or thermal decomposition or brittle destruction of BeC layers on the inner divertor (the JET study using QCMs cited above would indicate erosion of surface layers). There was some discussion as to whether this could put further restrictions on the tolerable ELM size for ITER because the radiation starts to have a major impact on the pedestal and subsequent ELM behaviour. In JET the enhanced radiation appears to force a temporary transition to a plasma with lower confinement - Type III ELM-like. II-2 - Long-range D retention, J. Roth Chair Long-range retention after D implantation was previously reported both in metallic and in carbon based plasma-facing materials. This session was devoted to document the status of understanding of the deep penetration far beyond the ion range. The session was divided into three talks dealing with carbon and carbide materials and three talks presenting results and modelling for the high-Z materials Mo and W. Dr. Tanabe summarized the experience from tokamaks for the case of carbon based materials and Prof. Kurnaev reviewed the existing knowledge on deuterium retention in CFC from ion implantation and plasma exposure studies. Depth profiles of retained D and T were measured by the analysis of cross sections of tiles on a mm scale, while NRA and SIMS analysis can

probe the first several µm into the material. The interpretation of such profiles implies a combination of direct ion implantation into the surface and gaseous penetration into the bulk along grain boundaries and absorption on the surface of individual grains. The assumption of hydrogen penetration through interconnected pores into the CFC was directly documented in the presentation by Dr. Roth. He reported on results from CEA on depth profiling of D in CFC exposed to Tore Supra using lateral and in-depth resolution by micro-beam NRA. It could be clearly shown that the D distribution was inhomogeneous both laterally and in-depth on a scale of µm. From correlation with TEM analysis of spots with high D concentration it could be concluded that hydrocarbons or hydrocarbon radicals penetrate through the porosity until the molecules are absorbed on inner surfaces. The second part of the session was devoted to deuterium retention in metals. Dr. Ogorodnikova presented experimental data for 200 eV D in W and temperatures up to 700 K together with detailed modelling of the total retention and the depth distribution. Numerical modelling of the ion implantation, surface recombination and release, diffusion and trapping in ion induced and intrinsic damage sites, shows retention in at least two different trapping sites associated with displacement damage and ion-induced extended defects in depths up to several µm. Modelling following a similar physical picture was presented by Dr. G. Wright for the case of D in Mo. He interpreted results obtained at the plasma device DIONISOS presented by Dr. Whyte. Similar to W, but with higher concentrations, the depth profiles show near surface retention and a diffusive tail into the bulk. The results of ion beam damage leading to increased sites in the lattice indicated that neutron damage in a nuclear environment could be very important for metals such as W or Mo leading to trapping site creation linear in fluence. The application of the modelling to the dynamic case of subsequent C-mod discharges could reproduce the linear increase in D retention with number of discharges, but at a factor of 10 lower level. This factor could be explained by D retention in a ≈ 10% B-intermixed surface layer on Mo. In general, the origin for deep penetration of D into CFC and metals has been elucidated. In CFC molecular migration through interconnected pathways and absorption on inner surfaces of pores is responsible, while for metals the irradiation produces extended damage structures well beyond the ion range which act as traps for diffusing hydrogen atoms. Of particular import for ITER was the fact that neutron damage will occur deep into the materials creating traps for T at a rate proportional to neutron fluence. Such retention could dominate over ion-induced surface damage for a number of reasons. For more detailed understanding and prediction for ITER more thorough experimental investigation is needed. II-3 – Joint session on ELM control, N. Oyama chair Dr. Fenstermacher reported recent results of ELM suppression with Resonant Magnetic Perturbation in DIII-D. A requirement of I-coil current was investigated using ITER similar shape plasma with pedestal electron collisionality ~ 0.1 from the view point of the width of the island overlap region. A preliminary analysis using vacuum fields and model profiles in the equilibrium reconstructions indicates that Δstoc

vac / Δpped > 3.5 is required for complete

ELM suppression, where Δstocvac is the width of the island overlap region including the n=3

RMP fields, n=1 field error correction and the remaining intrinsic field errors, and Δpped is the

width of the total pressure pedestal. It is also noted that the equilibrium reconstructed by Kinetic EFIT, which can treat edge bootstrap current consistent with pedestal pressure, is important to evaluate the width of the island overlap region, because the edge bootstrap current can change edge magnetic shear considerably.

Results of ELM mitigation by low n (1 or 2) external magnetic perturbation using Error Field Correction Coil in JET were reported by Dr. Liang. In a wider range of q95 (3.0-4.8) and ITER-like configurations, ELM frequency increases by factor up to ~4 and so that ΔWELM/WTOT reduces below 2%. Transport analyses shows no or only a moderate (up to 20% depending on operational scenario) degradation of energy confinement time during the ELM mitigation phase because of the density pump-out, but when normalized to the IPB98(y,2) scaling shows almost no reduction. The reduction in density and temperature at the edge is almost compensated by an increase of the central temperatures. The minimum perturbation field amplitude above which the ELMs were mitigated increased with decreasing q95 but always remained below the locked mode threshold. The attempts to reduce the ELM heat load using Error Field Correction Coil were also applied to MAST. Preliminary n=1 and n=2 experiments were reported by Dr. Kirk. In the case with n=1 perturbation, the delayed H-mode transition (or locked mode) was observed as coil current increased, even if Chirikov parameter exceeded unity at the pedestal region. In n=2 case, the location of the divertor strike point was shifted and the total amount of power to the divertor increased together with small density reduction was found in low collisionality edge (νe

*=0.1-0.3). But, no clear effect on ELM characteristics has been observed so far. Dr. Lang summarized current status of pellet pace making and prospect for ITER. In ASDEX upgrade, same pedestal parameters (ne and Te) as the gas fueled plasma with fELM of 51 Hz can be obtained with fELM of 83 Hz by pellet pacing. The result demonstrated that fELM can be controlled by external trigger of ELMs. So far, the controlled ELM frequency was about twice the natural ELM frequency in the previous experiments. But, ITER will require 5-10 times higher fELM than the natural ELM frequency, which will be demonstrated using new pellet injector in ASDEX Upgrade (143Hz) and JET (60Hz). Another remaining issue is the separation of ELM control from the fueling. In the discussion session, we have discussed some issues in ELM control using pellet pacing and Resonant Magnetic Perturbation. The way to optimize the pellet injector was discussed to achieve wide operational window in ITER and tangential injection was proposed to achieve shallow injection to minimize the fueling. The recovery of plasma density with edge ergodization was pointed out as an important issue in the demonstration of ELM mitigation with Resonant Magnetic Perturbation, because ITER will require some amount of density to achieve Q=10 It is also pointed out that edge ergodization combined with pellet fuelling should be explored on present machines. II-4 - Joint session on modelling (X. Bonnin chair) The session covered a wide range of modeling techniques, applied to a large number of machines. Topics went from interpretive modeling, using extensively the experimental data available, and doing a reconstruction of the plasma by means of the OSM-Eirene code. A similar approach can be taken with UEDGE. The insight and knowledge of plasma transport coefficients gained in such studies can then be leveraged for predictive simulations. A large amount of effort is also devoted to the more standard approach of predictive 2-D edge fluid plasma codes, often coupled to Monte-Carlo packages for neutrals. The issue of fuelling of the pedestal region, in both DIII-D and AUG, was tackled with the UEDGE code package and in Alcator C-Mod by neutral kinetic calculations. Matching the large SOL flows and ELM behaviour seen in experiment remains a challenge for such codes and is only achieved in special cases. However, significant progress on numerical methods for modeling drift cases with SOLPS5.0 was also presented and appears to yield those large flows more routinely.

Lastly, development of full 3-D turbulence codes, encompassing both the core edge region and a limiter SOL, is ongoing. The following discussion centered on the need to more efficiently leverage the ITPA database on pedestal and SOL profiles, which is currently sparsely populated with only a few DIII-D and AUG shots, so that code users may more easily validate their output against experimental results and/or participate in experimental proposals while experimentalists can compare the code results to their measurements. It was asked in particular that the Exhaust Task Force at JET make data available in the database format, although the request was not limited to JET only. It was also noted with some disappointment that few of the modeling efforts reported where aimed specifically at ITER issues but rather remained close to existing experiments at home institutions. In particular, the difficulty current codes have at reproducing all the features of divertor detachment, in particular the particle flux rollover, were seen to be of concern for proper extrapolation of current models to the ITER scenario of semi-detached divertor operation. II-6 - D retention (shot-integrated and campaign-averaged, chair E. Tsitrone The session was divided into 3 topics :

• recent results on fuel retention in tokamaks • retention in gaps • effect of conditioning

The discussion was focussed on comparison of fuel retention deduced from integrated particle balance and post mortem analysis, and on a first attempt towards multi machine scaling of retention. Progress in assessing the accuracy of the particle balance is acknowledged (comparison with active pumping on/off in TS, shots with closed valves in C-Mod, cryopump regeneration in JET). The comparison of the D inventory deduced from integrated particle balance (typically 10-25% of that injected during a discharge) and post-mortem analysis (typically 1-5%) are in marginal agreement (in AUG), given the error bars on both methods. To progress further, it is recommended to carry out more experiments whereby the amount of gas retained in vessel surfaces is deduced from direct pressure measurements after the discharge and compared to the amount of gas injected. This technique, perceived as more accurate than integrating the pumping speed x vessel pressures during the discharge, was applied to C-Mod and JET with talks at this meeting; C-Mod presented further details (some results at previous meetings) of single-discharge particle accounting with valves to pumps closed. JET recently developed the following scenario compatible with long periods required to regenerate cryoupumps to recover pumped gas: they repeated the same discharge throughout a day with pump valves closed and then regenerated the cryopumps afterwards. Other devices are starting to apply this technique (TCV, now equipped for particle balance). A dedicated experiment will take place in Tore Supra, with 2 weeks of repetitive long pulses, in order to cumulate in a short time a significant D inventory compared to previous campaigns, followed by an extensive phase of post mortem analysis for comparison with particle balance. Extrapolation to ITER from the present results is still difficult for the following reasons:

• Most present day machines are carbon dominated. Therefore, experiments on particle balance from fully-W AUG, C-Mod in Mo, and JET in the future ITER like wall configuration are considered as high priority.

• Most present day machines are running short discharges. As the retention is not seen to saturate in most devices (coherently with the codeposition or bulk diffusion

mechanisms) while the recovery after shot is shown to saturate at a constant value for large cumulated D inventory (whether for large gas injection and/or long shot duration), the retention fraction could reach higher values for pulse lengths considered in ITER.

• Most present day machines are operating with non-actively-cooled PFCs, and not in the temperature range foreseen for ITER. The first plays an important role in the dynamic behaviour of retention during the shot (outgassing from heating PFCs), the second could lead to lower D concentration in the materials in the case of ITER, at least on the divertor targets in the high temperature zone.

• In the reference semi-detached scenario for ITER, ELMs will probably be a major factor for material erosion, leading to short and intense high temperature plasma fluxes on PFCs on the otherwise low temperature background between ELMs. This is not the case in most present day experiments. Therefore, experiments in realistic ITER conditions (semi detached plasma) are recommended.

• It is recognised that the parameter often used to qualify retention, ie the retention fraction of the injected flux, although convenient, does not have an obvious physical meaning, and is therefore difficult to use for extrapolation.

Concerning the latter point, a first attempt is being set up to establish a multi-machine scaling of retention as a function of more physical parameters. Preliminary data have been provided by different devices (C-Mod, AUG, Tore Supra, Textor, JET, LHD). To progress further, it is proposed to :

• Establish a list of the useful parameters to implement in the scaling on the basis of the retention mechanisms (plasma parameters : particle flux to divertor/main wall surfaces, incident energy, fluence …, device parameters: operating temperature, PFC materials, PFC surface …)

• Use the existing database of retention in the different devices for a first try • In a later stage, propose coordinated experiments on the basis of this first attempt

II-7 - Next meeting & high priority tasks, Chair - B. Lipschultz The 10th ITPA SOL meeting is proposed to be held in Spain in January 2008 in conjunction with the PSI abstract selection meeting. We have asked A. Loarte to seek support from local labs for sponsoring the meeting. The 11th ITPA SOL meeting will likely be in Japan in the fall of 2008. N. Asakura will investigate the timing along with T. Tanabe. Possible locations include Kyoto (Mizuuchi sponsor?) and Kyushu (Tanabe sponsor). Next meeting topics:

1 - D/T issues 1a - Removal rates and applicability to various materials and hidden areas 1b - Tile side studies 1c - Tutorial on neutron damage and T retention 2 - Metals and mixed materials 3 - Limiter startup 4 - Further studies of SOL characteristics during outer limiter startup 5 - ELM & disruption heat loads 5a - limiter vs. divertor operation, 5b - radiation and time dependence, dust,

5c - mitigated scenarios, advanced scenarios 6 Divertor scenarios and detachment physics 7 - Dust 8 Erosion and heat loads due to fast/non-thermal/RF-accelerated particles 9 Tile design wrt. heat load and T retention 10 ITER diagnostics 11 Tutorial about hydrogen plasmas – what can we learn from the H phase in ITER?

II-8 - General modelling session, chair - X. Bonnin A large amount of effort is going on at several institutions across countries to (in)validate codes by extensive detailed benchmarking between numerical methods, codes and with experiments. This activity has grown from an initial impetus given by DSOL-14 but has now been extended to include contributors from Japan and comparisons with kinetic plasma codes. The level of detail required makes the work proceed slowly, usually at the pace of visits by code experts to each other’s institutions, under various auspices, for example in the EU by well-timed JET secondment agreements. However, the rewards of the activity are manifold. In addition to demonstrating a generally good level of agreement between codes (for relatively simple cases), thus improving confidence in their results, it brings to light discrepancies in implicit assumptions used in the physics of the codes, of which the code users are often unaware. For example, differences in classical transport formulation, treatment of heat flux limits, choice of atomic physics database, etc… can usually explain a large amount of the variance between different codes and can be readily addressed. However, it was also demonstrated that 2-D fluid codes systematically underestimate the radial electric field in SOL, a problem likely to be related to their mismatch of the SOL flows. Another large amount of activity is directed toward improving our understanding of the region of the far SOL and the wall. Several methods are being pursued to describe the far SOL region in terms of triangular meshes on which to follow either a fluid or a kinetic description of the far SOL plasma and neutrals, allowing inclusion of important physics for ITER such as neutral-neutral collisions, molecular physics, photon opacity, plasma transport during ELMs, and in front of antennae, etc… This work is done with coupling to the more traditional edge and SOL codes in mind. In parallel, developments of models for wall inventory of mixed materials, including eventually hydrogen isotopes, and their consequences on plasma scenarios are continuing. It was pointed out by the ITER Team that modeling needs will be shifting from scenario development to diagnostic design support as the period for procurement of the hardware moves on from the divertor itself to the diagnostics. More generally, the feeling of the community was that priority in the modeling effort should be given to understanding and reproducing detachment physics, which are an essential assumption for ITER operation, followed by work on kinetic and transient effects such as ELMs, and reproduction of the experimental signatures of radial electric field and strong Mach flows in the SOL. II-9 – ITER – K. Lackner

There was a consensus that ITER should have the capability for an (at least once in a lifetime) exchange of the first wall material. There was a general recognition that at some stage ITER should test a fully DEMO/reactor relevant material combination for both first walls and divertors, which in all likelihood appears to be tungsten. It would be highly desirable if the present design of the first wall could be modified in a way to give more flexibility for this, including the introduction of protection limiters. The majority of the group does not feel that presently available experimental evidence would justify now the start of ITER operation with all-tungsten plasma facing components.

There was an extended discussion on ITER plans for the hydrogen phase. The group felt that the period as presently allocated (3 years) was too short to properly get all diagnostics and control systems ready for the D/T phase when ITER will need to completely rely on remote handling to fix any problems. More importantly, there was a general consensus that the H-phase is the time to test our physics understanding and engineering hardware; hydrogenic species retention studies are needed to properly understand the level in ITER before the T phase. If the retention is too high then time will need to be taken to either modify the first wall materials or develop T removal techniques compatible with the existing materials and operation. The first-wall components need to experience heat loads (in particular off-normal ones) as close to that of the T-phase as possible. This translates into achieving H-mode as that increases the stored energy in the plasma for disruption and ELM heat loads. If the neutral beams are not available during the H phase one could imagine adding RF heating. Having higher energy transient heat loads during the H phase may also affect the T retention both through changed erosion levels and patterns, but also through the potential application of disruptive radiative heating of the PFC surfaces.

The status of the various ITER review committees was reviewed. There has been a very thorough review of the existing experience for conditioning, first-wall and divertor heat loads, and finally comparison of existing first-wall strategies with those currently planned for ITER. In all areas it appeared that important decisions needed to be made soon (e.g. structure of first-wall limiters, types of conditioning) as well as further research into heat loads and patterns from disruptions, ELMs and other off-normal events (e.g. the ITER requirement that the first-wall be able to handle 1 second of direct plasma contact).

Several presentations were given on predictions of ITER conditions. The first, on the allowable T limits in ITER, thoroughly reviewed all the possible T release mechanisms during accidental vents and other accidents. Radioactive decay heating of first wall surfaces is an important limit to T in the later phases of ITER - most of the T trapped in tiles will be released. The mobilization of dust is a concern, difficult to estimate, for all phases of ITER. In either case the current specification for T limits in ITER appears to be approximately correct given the uncertainties.

A separate study was made of how to extrapolate our current database of SOL characteristics to ITER with regard to parallel power flows. Relying solely on upstream SOL electron ne, Te profiles the data from a set of tokamaks gives a scaling with major radius (and some dependence on poloidal beta). This extrapolates to an ITER parallel power flow e-folding width of 6-8 mm. If true it implies that current tokamaks have similar, if not larger, parallel power flows as ITER would have. Strangely enough this result is also consistent with earlier divertor heat load profiles which, when mapped back to the midplane, give similar values. Further comparisons across tokamaks are needed.

II-10 - PFC material issues - A. Kallenbach This session covered various aspects of issues related to plasma facing components and analysis of surface analysis and spectroscopic data. S. Brezinsek reported that the photon efficiencies for various molecules do not change much during detachment in JET. As a consequence, the reduction of the intrinsic molecular photon fluxes during detachment can be interpreted as a reduction of the erosion. Flux dependence and energy threshold/dependence at low Te are the supposed origin. This behaviour is in contradiction to that reported from JT-60U (T. Nakano) in the previous meeting. In the

discussion, it was pointed out that the reason of this discrepancy, which could be related to the different surface temperatures, needs to be figured out. M. Mayer reported about the carbon erosion in the outer divertor of AUG. For the 2004-2005 campaign a total erosion of 2.8 g was obtained from marker probes, indirect evaluation using spectroscopy revealed 1 g. A strong reduction (fact 7) of C re-deposition in the inner divertor was found in campaign 2005/6, suggesting that the outboard limiters had been an important carbon source before coating with W.. G. Matthews reported about the ITER like wall project in JET, which will start with full W divertor. This helps to avoid contamination with C which could happen if the ITER like mix would be used initially. The JET shutdown is planned to start in Nov 2008. Currently, a huge effort on tile and remote handling tool development is performed. R. Neu reported about the current restart of AUG with full-tungsten PFCs. The machine has not been boronised to assess a fully high-Z wall. So far, breakdown and plasma current build-up could be demonstrated. Operation is planned for several weeks without boronisation to investigate, hydrogen retention, disruptions and ICRF in a C-free, all metal machine. G. Van Rooij reported about experiments in the Pilot Magnum device. Impressing plasma parameters could be achieved, and the reduction of chemical erosion at high temperatures has been investigated. A. Kreter reported about quartz-microbalance measurements in JET and their interpretation in terms of migration paths. The C deposition in inner and outer divertor louver during disruptions in JET increases with plasma energy. Remote areas show strong increases of particle fluxes, but low or no additional heat flux.. T. Tanabe reported about optical techniques for in-situ surface analysis. Tile analysis with Raman spectroscopy can yield valuable information about the structure of deposits. III- Session talk summaries III-I Session 1 Measurements of power deposition during ELMs in JET and ASDEX Upgrade – T. Eich For a physics based extrapolation of target power load characteristics it is necessary to understand the ELM related SOL transport physics. However, the striking observation that ELMs drive a larger fraction of the released pedestal energy towards the inner target plates than to the outer in normal field direction is not resolved. For this reason, dedicated discharges for optimized infra-red measurements have been performed in upper single null (USN) geometry with both normal and reversed field direction, i.e. with the ion BxgradB drift direction pointing towards and away from the active X-point, respectively. These measurements are complemented by an analysis of currents flowing through the inner and outer target plates. The experiments show that the ELM power load towards the inner target plate is larger compared to that towards the outer target with normal field direction but vice versa with reversed field. The current measurements also reveal that a net negative charge flows into the outer target and a net positive charge into the inner target during the ELM in normal field and also vice versa for discharges with reversed field. The absolute value of the ELM target energy difference between inner and outer target is strongly correlated with the corresponding charge difference in both targets. This strong correlation confirms the correct analysis of the target surface temperature when calculating the corresponding power fluxes to

the target tiles and is found to be consistent to recent studies at JET, where target current measurements are missing. It is conjectured, that the observed asymmetry of power/current is due to a net asymmetry of ions arriving at the divertor target plates. ELM-wall interaction – W. Fundamenski Type-I ELM-filaments were observed on JET in both visible (D-alpha) and infra-red (heat load); Type-III ELM filaments did not to leave a measurable IR imprint. In all cases analysed, ELM filaments were found to follow pre-ELM magnetic field lines, i.e. they did not noticeably distort/perturb the SOL (poloidal/toroidal) magnetic field. Toroidal separation Δφ ~ 2π /n, where n is the number of filaments (or the toroidal mode number in linear theory), was found as follows:

(a) in high triangularity, ITER-like shapes, n ~ 5 – 20 at the upper dump plate (~2 cm), and n ~ 20 – 50 at the outer limiter (>5 cm). This suggests ELM filament break-up in the far-SOL, consistent with ELM filament fine-structure observed on outer limiter.

(b) in low triangularity shapes, n ~ 10 – 15 was found at the outer limiter. The above toroidal mode numbers (n ~ 5 - 20) are in fair agreement with linear MHD (P-B model) of ELM instability. Despite large scatter in both the filament width and separation, both were observed to increase weakly with ELM size, i.e. n was found to decrease weakly (<linear) with ΔWELM; their ratio was independent of ELM size with an average normalised width of δ/Δ ~ δq/Δq ~ δf/Δf ~ 0.6 ± 0.2. This tendency predicts n >> 1 for small ELMs (eg.Type-III), and n → 1 for giant ELMs. Since the normalised upper gap, Δr/a ~ 2 %, is similar on both JET and ITER, one may expect similar behaviour in ITER. Divertor plasma-surface interaction and induced radiation under large ELM impact - R. Pitts Observations were reported from a recent experiment on JET designed to investigate the effect of large ELMs on the plasma-surface interaction and subsequently on the plasma itself. A series of vertical target discharges at 3.0 MA/3.0T (total injected energy up to 195 MJ) with varying fueling levels have been executed, producing Type I ELMs provoking drops in the stored energy in the range ~0.1 - 0.9 MJ (DWELM /Wped = 5-20%) and drops of 20-65% in the pedestal electron temperature. The ELMs in this series lead to an increase in the total radiation of ~50% of ΔWELM with considerable scatter in the range 25-100% but with a definite trend to lie above 50% for the highest ΔWELM. This applies both for the radiation induced by the initial ELM spike and to the total energy integrated over the compound phase which usually follows ELMs with larger ΔWELM. The strong “break” seen in the radiation seen in a similar discharge in 2003 have not been found in the new experiment. Large ELMs (obtained at zero fueling and often accompanied by long ELM-free phases) sometimes provoke a small increases (~10%) in Zeff derived from line integrated bremsstrahlung measurements. For ΔWELM up to ~0.6 MJ, the inboard-outboard divertor volume radiation asymmetry favours the inner divertor and increases from 2 to ~5, thereafter decreasing rapidly as the radiation in the X-point vicinity incurs toward the outboard region, but does not fall below ~2 at the highest ΔWELM. Infra-red thermography of the target tiles (data still to be validated) show surface temperatures not exceeding 800ºC on the outer target, but reaching ~2000ºC on the inboard target, due to the presence of surface layers (the ratio of conducted energies to the targets is ~3 in favour of the outer target) and the fact that ELMs convect energy preferentially to the inner target (see talk by T. Eich). These temperatures are too low for bulk target ablation on carbon, but may well be high enough to ablate the thin inner target layers, whose properties will be likely strongly dependent on history. Fast measurements (250 kHz) of ion saturation flux to Langmuir probes in the divertor and at the outboard first wall

limiters show an extremely rich structure, particularly in the compound phase, with wall ion fluxes which can be comparable (less than a factor10) with those arriving at the divertor. "Quartz microbalance (QMB) measurements of ELM induced material migration in the JET divertor" – A. Kreter In the recent JET campaign, the transport of carbon in the divertor was investigated by the quartz microbalance (QMB) technique. The results can be summarised as follows: (i) The transport is mainly line-of-sight, with particles predominately sputtered at the strike point (SP) positions and travelling over distances of several cm across the magnetic field; (ii) The amount of eroded carbon depends on the surface type, with lower rates for the bare CFC and higher for a-C:H layers. The highest rates are obtained after a shift of the SP position to the “fresh” layers deposited in previous discharges (“history” effect); (iii) The erosion rates in the inner divertor increase strongly non-linear with increasing energies of ELMs. Erosion induced by ELMs is attributed to the thermal decomposition of carbon layers in the inner divertor by carbon ablation or by increased production of carbon particle clusters (dust). In the outer divertor, where these thick layers are absent, no influence of ELMs is observed. Fast response of the divertor plasma and PWI at ELMs in JT-60U – N. Asakura Fast response of the divertor plasma and PWI at ELMs in JT-60U by Nakano&Asakura (1) Three He I , Dα and C II lines were measured at LFS diveror, and time evolution of plasma Te and ne (from He I lines, 50µs smpl.), ionization flux (from Dα line, 2µs smpl.), and C+-> C++ ionization flux (from C II, 50µs smpl.) were investigated during ELM. During ELM heat loading (~200µs), Te was increase. Then, Te was decreased due to ionization of recycled neutrals and reduction in heat flux. At the peak of Dα, carbon yield (GC+->2+/GD0->D+) became the lowest, suggesting chemical sputtering is increased due to large neutral sources. (2) ELM plasma transport at High Field Side SOL was further investigated: (2-1) Both ELM plasma flux and SOL plasmas at the far SOL were mostly transported from LFS. (2-2) Large multi-peak flux with fast flow (M//~1) towards the HFS divetor was measured only near separatrix (ΔRmid<0.4cm), which may caused fast plasma transport shorter than convection time from LFS (~180µs). (2-3) Impurity generation: both CII and Dα was increased simultaneously at HFS and LFS divertors. Neutral ionization flux was enhanced faster and larger than carbon ionization flux. III-2: Session 2 Depth-profiling and thermal desorption of hydrogen isotopes for JT-60U tiles – T. Tanabe Results of depth profiling and thermal desorption measurements of H, D and T in carbon tiles of JT-60U are presented and an isotope exchange mechanism and temperature effect for tritium release will be discussed. It is shown that hydrogen in the redeposited layers is easily exchanged with hydrogen isotopes succeedingly impinging. Accordingly, most of retained hydrogen in JT-60U tiles is protium because most of discharge campaigns are finished by HH discharges. The physics of long range retention in different materials – V. Kurnaev

Observations of long range penetration of hydrogen in vanadium alloys, WC, W2C and CFC are presented. Possible mechanisms of D deep retention are considered. H2 penetration through MPG-8 graphite is shown to correspond to gas transport trough capillary.

It was found that hydrogen implanted in V-3.49Ga alloy during irradiation by stationary, pulsed high-power plasma and 6-keV H+ ions diffuses through 0.7 mm sample and accumulates on its backside.

Irradiation of V-(10÷15)Ti-(6÷10)Cr-0,05Zr samples by a steady state hydrogen plasma led to increase of the microhardness through all the sample thickness possibly due compressive stresses created by hydrides formed.

Possible (but not proved) mechanisms of long range effects in polycrystalline metals are: enhanced diffusion of admixtures along interstitials and grain borders, the generation and transport of dislocation loops under radiation induced stress in surface layer and admixtures transport by shock waves.

Carbon precipitates are supposed to be trapping sites for D detected by NRA in WC and W2C up to 10 µm depth.

TDS measurements of D retention in NB31 irradiated at 500C in LENTA facility with 200 eV D plasma up to 1025m-2 gave the number of trapped particles two orders of magnitude higher than the ion stopping range can accumulate.

TDS measurements of D retention in N11 CFC from high field side tiles of Tore Supra for 1995 - 1999 period showed ~ 30 times higher retention as compared to MEPhI lab experiments. Possible reason is the difference of irradiation in tokamak, namely: presence of H in the CFC that increases 2 times D trapping at the expense of filling the H traps by D due to isotope exchange, plasma electron irradiation that increases D trapping 1.2 – 3 times, much more long range retention of D, more broad energy distribution of impinging particles, additional CFC damage and traps generation due to He GD conditioning, oxygen contribution to surface relief development.

H2 (and Ar) penetration through 1,3 mm thick fine grain MPG-8 graphite at room temperature is shown to correspond to molecular gas transport trough the capillary system of graphite with specific gas permeability 5·1015 mol·c-1m-1Pa-1. Sample made from T-10 MPG-8 limiter tile reveals 2 times larger permeability possibly due to increased porosity after long term operation in tokamak.

Lab experiments with well defined impact parameters are crucially necessary to qualify and predict the long range D retention in PFC. Effect of co-deposited layers the migration pattern of D into CFC – J. Roth

In ion and magnetron plasma irradiation studies large retention of D in CFC was reported without the saturation behaviour known from pyrolytic graphite. NRA ion beam analysis showed a high D concentration in the ion implantation zone with a tail into the bulk extending in depths larger than 14 µm. The diffusional tails revealed a diffusion with low thermal activation.

CFC materials have very rough surface topography and strong interconnected porosity as demonstrated by SEM and X-ray tomography analysis. Therefore, a new analysis technique has been applied consisting in micro-beam analysis, allowing a lateral and in-depth resolution of about 1 µm combined with SEM and TEM analysis. Samples exposed to the SOL plasma of Tore Supra were analysed, having a surface deposit of 3.5 µm a-C:D and a deuterium diffusion tail into the CFC substrate. While the D content in the deposited layer was rather homogeneous, the diffusion tail consisted in µm-size spots with high D concentration. A direct correlation of these spots with SEM and TEM images revealed the existence of large pores with deposited a-C:D layers on the inner surfaces, responsible for the high local D concentration.

It must be concluded that • D is stably retained in radiation defects within the ion range up to a concentration of

0.4 D/C;

• an excess amount of D recombines to D2 and is released; • hydrocarbon molecules and radicals are formed at end of range and migrate through

interconnected pores to the surface and into the bulk; • the migration proceeds with low activation energy until final, more stable trapping site

is reached. Co-deposition of deuterium on the inner surfaces of pores; • the amount of retained deuterium depends on implantation time and only weakly on

temperature. Modelling of ion-driven retention of D in W – Olga Ogorodnikova Modelling shows the presence of ion-induced and natural defects in polycrystalline W, which act as trap sites for deuterium. Ion-induced defects are produced during implantation by deuterium self-aggregation due to the stress field induced by the incident ion flux. The rate of ion-induced defect production depends on the energy of the incident ions, ion flux, sample temperature and initial trap concentration. Pre-implantation by deuterium ions with intermediate TDS increases deuterium retention. While annealing of W up to re-crystallization decreases D retention. Long-range retention in Molybdenum – D. Whyte Experiments were carried out on DIONISOS on D retention in bare Mo. The displacement damage caused by the 3.5 MeV He beam is found to enhance retention throughout the first 5 microns, raising concerns on the effect of neutron-induced displacements in ITER. Retention depth profiles indicate the beam-induced traps and plasma-flux induced traps are mobile at temperatures above 400 K. The retention profiles become constant versus depth at high T consistent with retention beyond the 5 micron detection limit. Dynamic temperature scans and cool-down show how permeated D can be frozen in deep into the Mo. A Dynamic Model for Deuterium Retention in Molybdenum. G. Wright A dynamic model for D retention in Mo has been developed based on experimental data from the DIONISOS experiment. The fits of the model to data from DIONISOS and Alcator C-Mod suggest a trap production mechanism in the implantation zone that scales linearly with plasma flux density. The model also shows that repeated thermal cycles with a rapid cooling phase allow the trapped D to progress deeper into the bulk with each subsequent thermal cycle. III-4: Session 4 - Joint session on modelling (X. Bonnin chair) UEDGE modeling of ASDEX-Upgrade and DIII-D – M. Groth UEDGE modeling of ASDEX-Upgrade and DIII-D low-density Ohmic plasma in normal BT configurations showed that the particle flux into the core is determined by a) neutral penetration both at the high-field and low-field x-point regions, and b) ion fueling at the top of the plasma via ion BxGrad B. Detailed validation of the UEDGE simulations were carried out against AUG and DIII-D pedestal (upstream) and divertor data. Ion flows in the main SOL were obtained in the direction of the inner plate for DIII-D when the radial diffusivities for the particles were increased by a factor of 4, and 4% neutral pumping was applied at the private flux wall. Yet, even with these assumptions the magnitude of the ion flow in the main SOL near the separatrix is below Mach=0.1, and the Mach profile increases with increasing distance from the separatrix toward the wall. Pedestal ionization profiles on MAST from OSM-Eirene – S. Lisgo

Simulations were carried out for MAST using the OSM-Eirene interpretive code package in an effort to quantify the radial and poloidal distribution of ionisation in the pedestal, for input to particle transport models of the H-mode edge transport barrier. Status of kinetic neutral modeling of pedestal fueling on Alcator C-Mod – J. Hughes Modeling of H-mode pedestal fueling on Alcator C-Mod has been performed, in order to gain greater understanding of the physical processes determining pedestal structure. Experimental pedestal data show the electron density pedestal having a linear dependence on plasma current, with cross-field transport increasing markedly as current is lowered. Varying target density alone has little effect on density gradient scale lengths in the edge transport barrier and a relatively weak impact on the height of the density pedestal. A modeled response of the density pedestal to perturbations to the edge neutral source couples a kinetic neutral treatment with a diffusive model for the plasma transport. The response is qualitatively consistent with the experimental pedestal response to target density variation, and the results emphasize the importance of pedestal screening in a regime of high neutral opacity. When modeling larger tokamaks operating at lower absolute densities, less substantial neutral screening is seen, and the extent of neutral penetration has a larger impact on the density gradient scale lengths. This modeling does not reproduce the typically clamped density pedestals seen in experiment during aggressive puffing into an established H-mode. Detailed analysis of edge profiles in these cases suggests that a simple diffusive model for plasma transport is deficient, and that a critical gradient assumption for transport may be essential for pedestal modeling. More advanced edge modeling of this kind is called for, including 2D modeling of the neutral source distribution and ionization profiles. Continued improvements to the C-Mod edge diagnostics set make advanced interpretive modeling of the edge very attractive. Modelling of Impurity seeded JET & ITER Discharges R.Zagórski and G.Telesca Impurity seeded JET and ITER discharges are numerically simulated by coupling a 1D multi-fluid model for the plasma core with a 2D model for the SOL-divertor region. The model is fully self-consistent with respect both to the effects of impurities on the a-power level and to the interaction between seeded and intrinsic impurities. This interaction leads to a significant change in the intrinsic impurity fluxes, and it is found to be essential for a correct evaluation of the average power to the target plates. The code has been successfully benchmarked against JET experimental data from the type III ELMy H-mode nitrogen seeded discharges. Extrapolation of highly radiating JET discharges to ITER indicates good performance of ITER plasmas with Q above 10 and large radiation (> 75 %). Simulation results show also that accumulation of helium in ITER might not be a serious problem. New version of B2SOLPS5.0 and simulations of H-regimes – V. Rozhansky A new version of B2SOLPS5.0 code has been developed and implemented for the simulation of H-mode shots. A new equation system is suggested, which is equivalent to the system which is used in B2SOLPS5.0 at present. The main idea is to replace large radial B! driven convective fluxes by parallel fluxes with the same divergence both in the particle balance and energy balance equations. This is of special importance for H-mode where the diffusion coefficient is strongly reduced inside the barrier. H-mode shot of ASDEX-Upgrade has been simulated with the new version with reasonable time steps and convergence. Interpretive transport simulations of the edge region with UEDGE – T. Rognlien UEDGE has been generalized to run in an interpretive mode for the pedestal region to determine transport coefficients there while simultaneously simulating the SOL plasma. New

2d2v kinetic edge simulations from TEMPEST without Er show much larger particle and power flux to the inner divertor than outer for the ion grad_B drift toward the X-point. Such Er=0 simulations might roughly correspond to conditions when an ELM substantially reduces Er in experiments and shows more ELM power to the inner divertor. Overview of "plasma reconstruction" modeling – S. Lisgo Dedicated interpretive modeling has the potential to contribute to the understanding of many plasma transport issues that are of critical concern to ITER. Efforts to generalize the modelling tool set are presented, such as extensions to 3D, a fluid-kinetic model, and time dependence. Presentation of the Tokam-3D code – A. Grosman TOKAM-3D, a new generation edge code for transport and turbulence is introduced. This code solves fluid drift equations in 3D full torus geometry, taking coherently into account drifts and curvature effects at all scales. The simulated area covers both closed and open field lines across the LCFS. First simulations are used to address the issue of parallel flow asymmetries and exhibit results in qualitative agreement with experimental trends, although it models now only an isothermal plasma. Two different mechanisms are evidenced, one linked to the strong ballooning of the radial turbulent flux, the other linked to large scale drift flows and their interaction with curvature. III-6: Session 6 - D retention (shot-integrated and campaign-averaged, chair E. Tsitrone) D retention (Shot-integrated & post-mortem) with Mo PFCs in C-Mod – B. Lipschultz Post-campaign analysis of tiles shows D/Mo levels below 1% to the depth that NRA measurements allow – 5 microns. We can then estimate the D retention rate averaged over the previous run campaign assuming that the D concentrations quoted reach 100 microns into the tiles. The resultant overall D retention is under 1% of the gas injected over the run period, < .01% of the ion flux to the outer divertor. Shot-integrated D retention was measured using 2 methods; pump gate valves closed during the entire discharge and for 5 minutes afterwards; and 2) calculation of what is pumped during the discharge and for 10 minutes afterwards based on pumping speeds. The amount injected into a discharge is measured utilizing calibrated plena and gauges. The latter is much less accurate. Results show that for discharges without disruptions the retention is often in the range 30-50% of that injected (2-5% of the ion flux to the outer divertor). The retention is highest when the divertor is in a high-recycling regime and the plasma is diverted. The retention appears to not saturate and increase linearly with pulse length. When one integrates the measured discharge retention over periods of days or weeks one finds that often disruptions dominate and reduce retention to much lower levels, if not reversing to outgassing. Given the variability of disruptions the shot-integrated D retention, while high in individual discharges, is consistent with the quite low campaign-integrated estimate. D retention (Shot-integrated & post-mortem) with C PFCs in JET – V. Philipps Dedicated gas balance sessions have been performed at JET under L mode , type III and Type I plasma conditions based on cryopump regeneration before and after (typically 2 hours) the experimental sessions. The integrated amount of fuel injection reached 4.5, 4.92 and 2.4g D respectively with a total amount of retained fuel (until cryopump regeneration) of 0.56 (L mode), 0.8g (typeIII) and 0.4g D. This correspond to a fuel retention rate of 2.09x1021D/s

(1.34x 1021 D/s ), 1.31x1021 D/s (0.82x 1021 D/s ) and 2.66x1021D/s (1 x1021 D/s ) when normalised to the overall external divertor heating time (or the total divertor operation time). This must be compared with a retention rate of 4.1 x 1020 D/s (2.6 1020 D/s ) evaluated from campaign averaged post mortem tile analysis in the last MKIISRP operation campaign. The accuracy of the JET gas injection and regeneration had been checked using the same injection systems and regeneration in dry runs ( no plasma) reaching a fuel injection of 3.18 g and a regeneration of 3.16 g D, showing that the above retention data are not determined by data inaccuracies. The data show that the amount of retention is not strongly dependent on the operation mode. The higher retention compared with post mortem analysis must be explained by the contribution from long term outgassing , the fuel release by plasma operation and wall conditioning. The data confirm again the strong concern with long term tritium retention in a full carbon wall environment. D inventory in ASDEX-Upgrade – M. Mayer Erosion/redeposition patterns as well as deuterium inventory determined by post mortem analysis were presented for the 2002-2003 and 2004-2005 campaigns in Asdex Upgrade, when the machine was still dominated by carbon. The inner divertor is deposition dominated while the outer divertor shows net erosion. For the 2002-2003 campaign, the major fraction of retained D (60%) is trapped on inner divertor tiles, while 20% is found in remote areas below the roof baffle, and the rest is distributed in tile gaps, on upper divertor tiles, on limiter tiles etc. The D inventory in gaps is small, and is found to change the spatial distribution of the retained D rather than increasing the total inventory. The D inventory in pump ducts is negligible. The long-term retention deduced from this post mortem analysis is 3-4% of the deuterium input, to be compared with 10-20 % from integrated particle balance. D retention in impurity seeded long pulse Tore Supra discharges – E. Tsitrone In order to study fuel retention at lower Te, a long pulse scenario with a feed back of impurity injection on the radiated fraction was developed. Radiated fraction from 30 (no impurity injection) up to 80 % were obtained, corresponding to Te in the range 30-50 eV at the last closed flux surface. The tested impurities (Ne and Ar) are mostly exhausted by the active pumping system. The particle balance for deuterium shows a decreasing retention rate with increasing radiated fraction, both in absolute and relative (compared to D injection or recycling) values. Analysis is still underway to explain this behaviour (potential explanation : outgassing from the upper part of the machine heated up by radiation, change in the carbon erosion/redeposition behaviour, plugging of porosity …). Investigation of dynamic retention under near-saturated wall condition - T. Nakano The relation between the divertor- and the wall-pumping rate observed in JT-60U long-pulse discharges was described. A high divertor-pumping rate results in a low wall-pumping rate. On the contrary, a low divertor-pumping rate results in a high wall-pumping rate. This seems to indicate an equilibrium between the divertor- and the wall-pumping rate. Thus, the divertor pumping could control the dynamically retained particles in the wall. Overview of Long-term Fuel Inventory and Co-deposition in Castellated Beryllium Limiters at JET - M.J. Rubel and J.P. Coad

This contribution provides an account of the detailed examination of several castellated beryllium tiles from the belt limiter (of which approx. 2000 pieces were in total) exposed to the JET plasma for 56000 s. The major aim of the investigation was to determine

the fuel retention and material mixing on the tiles. Analyses have been performed on both sides of castellated grooves, on plasma-facing and side surfaces of the tiles. The essential results are summarised by the following: • deuterium retention in the grooves is associated with co-deposition of carbon • decay length of deposition in the castellation is short: l ~1.5 mm; • the maximum deuterium content in the groove does not exceed 8x1017 cm-2; this is

measured at the distance of approximately 2 mm from the entrance to the gap; • no deuterium (above the detection limit of NRA technique) is detected in bulk Be; • bridging of gaps by molten beryllium is observed, but gaps are not filled with Be; • fuel content on plasma-facing surfaces is fairly low (in the range 1-8 x 1017 cm-2); it is

mainly associated with carbon co-deposition • on side surfaces of the tiles the formation of BeO layer is detected at a distance of 20 mm

and more from the plasma-facing surface. On carbon deposition in the first wall gaps – A. Kukushkin A qualitative description of the phenomenon of preferential carbon deposition in the first wall gaps was proposed. According to the model presented, the major effect causing this deposition is the flux dependence of the chemical sputtering yield: the higher flux density of the hydrogenic ions on the leading edge of the gap (as compared to the tile faces) makes re-erosion of the deposited carbon there less efficient, thus shifting the balance towards net deposition. In combination with reflection of atomic hydrogen from the wall, this model predicts non-monotonic profiles of carbon deposition down the gaps. For the tritium co-deposition with carbon in the gaps in the first wall in ITER, this effect yields ~ 0.2 g/pulse as the upper estimate. Effects of ICRF conditioning on the first wall in LHD – N. Ashikawa Using material probes located at the bottom of the vessel in LHD, the area affected by ICRH conditioning was estimated. Only plasma facing surfaces show damages with helium bubbles, while the other surfaces have few or no damage. This result suggests that ICRF wall conditioning is difficult for shadowed areas. In the future, the same experiment is planned in EAST. During ICRF wall conditioning experiments at different harmonics, high energy distributions were measured using a Si-FNA (fast neutron analyzer). At the 2nd harmonic mode for hydrogen (1.5T, 38MHz) no high energy distribution was found, while the fundamental mode for hydrogen (2.75T, 38MHz) exhibits a high energy distributions. The hydrogen removal rate is higher for modes with a high energy tail. During ICC, particle flux was measured by NDD (Natural Diamond Detector) to be about 1.6-3.2 x10^5. Wall conditionings for hydrogen removal on metal walls in EAST – J. Hu

ICRF wall conditioning with a specially designed antenna, including He-RF and He/O-RF (Ratio of He to O2 was 1:1), was successfully carried out in EAST, a divertor superconducting machine with metal walls. Much better results have been obtained compared with those in HT-7.The removal efficiency for H in low pressure (20kW 4.5E-3Pa) He-ICR is almost as the same as in He-GDC cleaning (2Pa). Factors influencing the He-ICR efficiency, such as magnetic field, RF wave duty time, power, pressure were investigated. In EAST, breakdown pressure for He-RF cleanings could reach 10 Pa and Highest H removal rate is higher than that in HT-7 by a factor of 4. He/O-RF on a stainless steel wall in EAST are beneficial for both H and C removal, similar as that on C walls in HT-7. Highest removal rate are 7.8×1022H-atoms/h,4.2×1022C-atoms/h

(20kW 7×10-2Pa), which were higher than that in He-ICR by a factor of 5 and a few tens respectively. Behaviour and influence of power, pressure and wall conditions were similar as that in HT-7. Removal rates in EAST were higher than in HT-7. O consumption rates (min-

1.m2) were lower than HT-7 by a factor of >10. III-7: Session 7 - Next meeting & high priority tasks, Chair - B. Lipschultz See summary section. III-8: Session 8 - General modelling session, chair - X. Bonnin Comparisons of SOLPS, UEDGE and EDGE2D – D. Coster The edge-code benchmarking activity has continued. For the D only, no-drifts cases satisfactory agreement between the fluid plasma, kinetic neutrals code has been achieved, and the D+C, no drifts cases are in progress. For the fluid plasma, fluid neutrals codes, the agreement for the D-only, no-drift cases, the agreement is much improved. Amongst the physics issues causing substantial differences, the most important have been: parallel heat flux limits, neutral flux limits, Braginskii vs Balescu formulations and atomic physics. SOLPS5.0 and SOLDOR/NEUT2D modeling of JT-60U discharge – H. Kawashima Benchmark between SOLPS5.0 and SOLDOR/NEUT2D codes have been carried out using a L-mode discharge of JT-60U tokamak. Initial results show that simulations of 2D profiles of electron temperature and density, and neutral density almost agree with both codes. We will adjust the treatment of impurity radiation in both codes because there was slight difference of radiation power. Simulating detachment in ASDEX Upgrade, TCV, and JET – M. Wischmeier The principal aim is to validate the SOLPS5.0 code, using the fluid code B2.5 coupled with the Monte Carlo neutrals code EIRENE'99, versus experimental data from TCV, ASDEX Upgrade and JET. The particular focus is on divertor detachment in order to assess the quantitative importance of the different processes able to determine the detachment threshold. The experimental data used are from as simple as possible ohmic and L-mode discharges, of which some have been done with He as a main fuelling species (TCV, JET). It can be assessed that currently SOLPS5.0 is in principle able to simulate detachment and the 'roll over' of the ion target flux. However, detachment as observed in experiment (D fuelled discharges in JET MarkII GB, ASDEX Upgrade DivIIb), in particular the detailed in-out asymmetry observed in forward field, are not reproduced (e.g. complete detachment at the inner target at relatively low midplane separatrix densities). Nevertheless it is possible, following a detailed cmparison of experimental observations and numerical results, to understand quantitatively the basic mechanisms leading to detachment, as shown in particular for He fuelled discharges at JET and TCV. BIT1-SOLPS comparisons – F. Subba 1D ELMs fluid models benchmarking against kinetic results is actively going on. Fluid and kinetic models show fair agreement for a pre-ELM steady-state phase. The onset of a numerical instability in the fluid model is currently preventing a detailed comparison with time-dependent kinetic results. Numerical experiments showed that the problem is not

determined by a too large time step, which is set smaller than the electron free-flight time in the system. Various strategies to improve the convergence of the fluid model are being investigated. Fluid modeling of the far SOL region – F. Subba Development of the ASPOEL code with the aim of coupling with the SOLPS tool is going on taking an ASDEX Upgrade shot as a reference case. A mesh has been produced for the far-SOL region, matching the outer wall and, on the inner side, a corresponding SOLPS mesh. The interfacing routines to pass information from SOLPS into ASPOEL as input parameters are being developed and tested. Discrepancies between modeling and experiment – A. Chankin Detailed SOLPS modelling of well-documented AUG plasmas reveals a tendency for code solutions to predict colder and denser plasma in divertor (target Te profiles flatter than in exp., peak Te smaller). Simulated Er in the SOL (SOLPS, EDGE2D) is much smaller than in experiment. The code underestimate of Er is consistent with their underestimate of experimental parallel SOL ion flows. Our hypothesis is that divertor, Er and flow discrepancies are all related to each other and caused by non-local effects of parallel electron transport in the SOL and divertor Improvements to the wall model in B2.5: coatings and layers – X. Bonnin The self-consistent modelling of plasma edge behaviour and plasma-wall interactions is becoming more and more a necessary ingredient for the ITER design. Efforts have been made to generalize an already existing module within the SOLPS code suite to treat PWI while also taking into account redeposited layers, their temperature and composition evolution, etc... Additionally, in order to be able to treat high-Z materials such as tungsten, a bundled charge state model has been implemented in the code. Transfer of SOLPS4.2 ITER simulations to SOLPS5.1- X. Bonnin Most ITER plasma edge simulations have been done with the SOLPS4.2 code, which couples the B2 fluid plasma package to the latest Eirene Monte-Carlo neutrals code. Some work has begun to migrate these runs to SOLPS5.1, which uses instead the more recent B2.5 package, including drifts and currents as well as better numerical treatments. The current status of this work and remaining hurdles were presented. New physics options in the B2-Eirene modeling package for ITER - V. Kotov It has been shown that taking into account the neutral-neutral collisions, opacity of Ly radiation and detailed description of the hydrogen molecular kinetics is important for modelling of the ITER divertor plasma. All those effects are now taken into account in SOLPS4.2 code. Molecular kinetics has been found to be important for high density and high power JET discharges as well. However, in the latter case including those effects does not help to eliminate all the disagreements between the modelling and experiment. Current status on ITER divertor modeling – A. Kukushkin Added to discussion. III-9 – ITER, chair – K. Lackner ITER Wall conditioning design review committee report – C. Grisolia The Glow Discharge system appears to be efficient:

• The remaining field created by the ferritic insert is lower than 10 gauss and is not perturbing the glow plasma

• The Anode design could be easily improved (several improvement will be proposed) However, several open issues remain unsolved:

• The anode seems to be too close to the divertor and there is a risk of a poor conditioning of the vessel walls.

• The pumping speed currently considered seems to be small for a good conditioning efficiency

During this Glow Design review, it was proposed to study the installation of fixed anodes which could be placed on the wall as done in several machines (Tore Supra, JET, AUG,…). Moreover, results from an ongoing study dedicated to push the glow plasma onto remote surfaces (ports, under the divertor dome) has shown that this could be achieved with secondary electron emitters. These two improvements could be part of an ITER design modification. To recover from heavy disruption, ITER ICH system appears to be the only reliable system to be used for conditioning during operation. Due to lack of budget, no dedicated antenna could be considered and the ICH system must be used. However, in order to establish the good ITER ICH recipe and to develop and test antenna protection during ICH conditioning operation, a comprehensive Work Programme must be undertaken in current machines. This WP will be proposed by the ITER WC working Group. Plasma treatment could be also considered in order to recover tritium trapped in Plasma Facing Components or on remote areas (as the part under the divertor dome). Controlled disruption or parasitic plasma as observed in AUG must be studied in a comprehensive Work Program involving several machines worldwide.

The use of oxygen in Tokamak to recover tritium trapped in deposited layers has been discussed. It seems that oxygen will not work in the ITER Vacuum Vessel due to too low wall temperature and ITER mixed materials. Furthermore, oxygen in ICH plasma could lead to flaking and an increase of dust inventory. In that case, it has to be pointed that only plasma wetted surfaces could be treated by the ICH oxygen plasma. As a conclusion, it is proposed to create a forum on this special open issue on the use of oxygen in Tokamak. This working group will coordinate a “systemic” approach of the subject, propose a work program to be undertaken in current machines and optimise the parameter which could be foreseen during ICH conditioning. ITER first-wall strategy design review committee report – P. Stangeby Some major issues: (a) wall loads (covered by Alberto Loarte’s presentation), (b) ability to change the main wall quickly: Remote Handling capability, (c) viability of the port start-up limiters. Regarding (b), from "Current ITER remote handling equipment requirements and availability for the exchange of the ITER first wall", by A.Tesini, Remote Handling Section, Tokamak Department, ITER IO, presented to Meeting of WG8, 16-20 April 2007: "The baseline ITER RH equipment is currently adequate to meet the requirement for the exchange of a few (≤ 20) blanket modules......Full blanket replacement (440 modules) would have a huge impact on the Hot Design. The machine components’ users are strongly encouraged to consider this limitation and to apply pressure on the Project if they think this is not a satisfactory situation." Regarding (c), conclusions regarding the port start-up limiters: (a) the start-up scenario needs to be re-assessed to definitively establish the maximum power to these limiters since their viability may be marginal regarding deposited power density, (b) the 4 mm

Be layer on the front of the limiters may erode rapidly; the ability of a thinner (eroded) Be layer to handle power needs to be assessed re the maximum permitted Be/Cu temperature (700C). (c) Zeff may be ~ 4; is this acceptable for the start-up plasma? ITER Design Review process; two steps; first step: establish if the present design is adequate. Report due by end of May. If conclusion is that design is not adequate, then second step: produce a re-design solution. Conceptual design due by September. ITER PFC heat loads design review committee report – A. Loarte The specifications of steady-state and transient loads assumed in ITER show some discrepancies with respect to those obtained by physics-based extrapolation from present devices in terms of prescribed values, definitions or absence of specifications. In particular : a) plasma particle and power fluxes along the field onto the first wall in steady-state conditions, including transients (~ seconds timescale) due to inaccuracies of the plasma position and shape control are not specified and can lead to significant loads on main wall PFCs (particularly at exposed edges), b) the power flux on the limiter during ramp-up is not specified (only the total amount of power) and no specifications are given for the ramp-down limiter phase and modelling shows that the power fluxes maybe larger than the design value for this component, c) ELM power loads on the main wall do not relation between ELM loads and ELM total energy loss nor their spatial distribution nor any in/out asymmetry at the divertor, d) disruption power and mechanical loads are only specified for the high performance phase of scenarios 1,2 and, possibly, 3 with the implicit assumption that these provide largest loads on any PFC than scenario 4 and the other three assessed scenarios, which may be invalid, e) specifications for disruption thermal and mechanical loads during the limiter ramp-up and ramp-down phases are missing, f) specifications for thermal loads on PFCs during VDEs and MARFEs are oversimplified (energy or power flux to the wall and timescales are only prescribed) and can over/under estimate the power fluxes to the PFCs during these events in ITER, h) specifications of radiative loads during the current quench of disruptions are oversimplified and are not provided for mitigated disruptions, i) specifications for runaway loads on PFCs during disruptions are oversimplified (energy fluxes to the wall and average timescales are only prescribed), which can underestimate the peak fluxes to the PFCs during these events in ITER and j) the values of local power fluxes on PFCs due to heating systems are either not specified (ICRH, NBI first-orbit losses, etc.) or oversimplified (i.e. for shine-through only average flux on inner-wall with no impact angle given), which can lead to significant loads and erosion of PFCs. The in-vessel component design review working group is launching and series of R&D and documentation activities to produce and new set of updated specifications of these loads, which will be used to evaluate the ability of the present PFC design to meet the requirements for ITER operation associated with these loads and explore alternative options. Scalings of q_parallel to ITER – D. Whyte The ITPA divertor/SOL database was analyzed across different devices using an upgraded sepatarix locator based on parallel power balance. The Te gradient scale length is found do vary linearly with R and be insensitive to other parameters including exhaust power into the SOL. This suggests the separatrix exists at a critical gradient set by curvature. The resulting upstream q// levels are ~ GW/m2 for present devices, commensurate with that predicted for ITER. Suggestions were made for more accurate upstream exhaust measurements, particularly in the ion channel. Physics assumptions that determine the T limit in ITER – P. Stangeby

Are the physics assumptions underlying the present T-limit correct? How much of the T in

carbon codeposits is actually mobilizable in the event of an accidental vent to air? Same question for T in CFCs? Same for T in Be or BeO or Be/C codeposits? Same for T bred in Be? Same for T in W? Is this the relevant accident scenario? water/steam leak? both simultaneously? dust? Is early-phase ITER (no radioactive decay heat) the same situation as ITER with 10 MW of radioactive decay heat? Is it possible to reliably predict this complex situation or will experimental simulations in ITER, both in pre-DT and DT phases, be needed?

It appears that, in the event of an accidental air vent occurring in the early phases of ITER, before radioactive decay heat has become high, only a fraction of the tritium contained in carbon codeposits that remain in the vessel will be mobilized because of rapid cooling of tiles. Release of T in tritiated dust may then be the controlling process. The problems of dust control/reduction is a more general one and must be solved in any case. In later-phase ITER, when radioactive decay heat has built up to MW levels, all in vessel tritium – whether in C, Be or W - should perhaps be considered to be mobilizable in the event of a combined accidental air vent and loss of cooling accident. Operational sequence of ITER for H, D, and DT phases – V. Chuyanov

The ITER project is starting now the procurement process. Long lead components- magnets , the vacuum vessel must be procured first. The procurement of in vessel components and external systems may be started somewhat later (depending on the component)There is still some but very limited time to incorporate in the design latest results of the fusion R & D.

The most difficult and controversial issue is the selection of materials for plasma facing components – the first wall and diverter.

The current ITER baseline selects: beryllium as a material for the first wall; CFC for strike points of the Diverter, and tungsten for the dome and upper parts of diverter plates.

The choice is an obvious compromise between different requirements and is a target of a strong criticism. There are several other suggestions like an all carbon machine, an all tungsten machine, or all possible combinations.

The very existence of so different proposals shows that now there is no obvious best choice. At the same time the project logic dictates the necessity to establish a clear time table of project decisions which has to take into account the time necessary for R@D, technological development, design, procurement, production and installation of different plasma facing components.

A logic and a time table of decision making process compatible with the ITER project schedule were presented. III-10 - PFC material issues, chair A. Kallenbach Hydrocarbon spectroscopy in JET (mainly the effect of detachment) – S. Brezinsek High density L-mode discharges in JET with outer divertor detachment have shown a reduction of the CD A-X band light which is an indicator of chemical erosion of carbon-based plasma-facing materials. In order to distinguish if the reduction is attributed to a change of the measured CD photon flux or the deduced Cd particle flux an in-situ calibration with hydrocarbons was performed. CD4 was injected through GIM 14, a single valve in the LBSRP, into the near scrape-off layer of the outer divertor. The target plasma was a high density L-mode plasma with attachment-detachment oscillation. The injected amount of molecules leads to a comparable emission of CD photons in both phases indicating comparable photon efficiencies in recombining and ionising plasmas. This gives evidence to a clear reduction of the chemical erosion of CFC in the phase with (semi-) detached outer

divertor plasma. Exact quantification is subject of calibration of the line-of-sight of the spectroscopic sytems with view on the single gas inlet location. Carbon erosion and transport in ASDEX Upgrade – M. Mayer Carbon erosion and transport in ASDEX Upgrade were determined during several discharge campaigns between the years 2002 and 2006. The step-wise increase of tungsten coverage allowed the identification of net carbon erosion areas. Between 1 g (spectroscopy) and 3 g (post-mortem analysis) carbon are eroded in the outer divertor during 3000 s plasma discharge time, while 14 - 15 g C are deposited in the inner divertor. The majority of the deposited carbon originated from the ICRH protection limiters at the low field side, and the inner divertor carbon deposition decreased to about 2 g after coating the ICRH limiters with W. Boron in the the inner divertor originates from erosion of boronisation layers and is about 3 - 6 g. Open issues and programme implications of the ITER-like wall experiment – G. Matthews The ITER-like Wall Project (ILW) project is part of a programme of JET enhancements (EP2) to be implemented before and during the Framework Programme 7. The ILW project was the subject of an international workshop held in October 2004 and was approved by the STAC AHG and given top priority within EP2 in February 2005 with the final go ahead given by the EFDA Steering Committee in April 2005. After this, the ILW team was built up and R&D tasks launched. The project is lead by JOC with core engineering and installation related activities based at JET including secondees from other EURATOM Fusion Associations. The R&D tasks have been based outside JET but under the overall management of the JET Operating Contract (JOC) ILW team. The current objective of the ITER-like Wall (ILW) Project is to install in JET, a beryllium (Be) wall and an all tungsten (W) divertor (the favoured back-up materials solution for ITER). This choice is technically more demanding than the ITER-reference combination which has CFC tiles at the strike point and leaves open the option to remotely replace specific rows of divertor tiles with CFC at a future date. The ILW will provide a test bed for integrated scenarios with ITER relevant edge conditions and compatibility with the wall, thus speeding up the early phases of ITER. The main constraints on the ILW project are cost, schedule and the need to preserve the power, energy handling and force limits (due to disruptions) currently set for the CFC wall whilst providing a scientifically relevant materials configuration for ITER. These constraints have led to Be tile designs which are inertially cooled (cost/complexity), segmented (eddy forces) and castellated (thermal stress cracking) with hidden bolts (maximum power handling). There has been great attention given to the tile geometry with the objective of avoiding field line penetration which could expose edges to a depth of more than 40 microns for a wide range of normal and off-normal plasma configurations. As well as idealised analytic local calculations, ILW are working with CEA to model the shadowing between multiple components meshed directly from CATIA files and to explore the effects of tolerances. This level of design is more stringent than has yet been applied to ITER where the wall load is considered to be radiative (apart from the limiters), but is necessary if the power handling due to plasma contact is to be maximised.

The recessed areas of the inner wall are either Be coated inconel (cost/eddy forces) or W-coated CFC (NBI shine-through). For the divertor, R&D tasks have been completed to evaluate W coating of CFC tiles (driven by cost/schedule) and the design of a single row of

bulk W tiles for the outer divertor. To meet the schedule, a major upgrade of remote handling systems and new tooling are required to install the novel tile designs. The existing short remote handling boom is being extended to allow co-operative working, with the new boom delivering complete kits of tiles and tools to the place of work . The ILW project met its objective to commit the material contracts for Be and CFC in 2005 (call for tender issued June 2005). Defining the blanks for >4000 main chamber tiles on this timescale was particularly challenging. Establishing the feasibility for recycling of 4 tons of tritium contaminated beryllium tiles left over from previous JET operations was a key factor in keeping the material procurement close to budget while eliminating a potential waste liability. This contract is almost complete with most of the material delivered to JET. Analysis of 3µm thick W marker stripes from the previous JET campaign suggests that thin coatings (10µm) have a marginal erosion lifetime for the outer strike point. High erosion was also seen on the horizontal section of the divertor baffle so a 200µm VPS coating and a bulk W tile row are proposed for the outer divertor. The inner divertor is a strong deposition zone and the 3µm W stripes appeared intact under a thick layer of Be and C. However, due to concerns about Be/W alloying a 200µm VPS coating is proposed here. In the recessed areas of the main chamber (e.g. NBI shine-through protection), 10µm W coatings are expected to have sufficient erosion lifetime due to sputtering and a lower disruption risk in the event of flaking. Although the ITER-like Wall will not be available for experiments until early 2010, it is important to define any special requirements now, including reference pulses which would need to be executed in the next campaign. Restarting JET with the new wall will be very similar to starting ITER and ITPA should consider which issues need to be addressed at each phase of the programme. Specific requirements such as recovering dust for analysis, analysis of D content of bulk materials etc. will need to be considered and scheduled - input from ITPA would be valuable in identifying and defining what might be done and linking it to experiments in other devices. The ITER-like Wall Project represents a significant step by EFDA in preparing for ITER operation. It is already shaping the programmes of the JET Task Forces by making the severe materials driven operational constraints expected for ITER a milder but still challenging and more immediate reality for JET. Initial AUG operation with 100 % tungsten PFCs – R. Neu During the last vent ASDEX Upgrade has exchanged all remaining graphite PFCs with tungsten coated ones. The last major component to be exchanged was the lower divertor, which now is equipped with a 200 µm W VPS W coating at outer strike-point region and a 4 µm PVD at the inner strike-point. Operation of ASDEX Upgrade has been resumed at the end of April. Due to the damage of one out of three flywheel generators last year, major reconfigurations of the power supplies were necessary, which have a strong impact on the plasma operation, making the restart procedure more challenging. Despite this caveat it was decided that the start-up will be performed without prior boronisation to allow for a full exploitation of the 100% W wall. Plasmas with a short flattop phase and auxiliary heating of up two 5 MW were achieved till the date of the ITPA meeting. The total radiation in the plasmas was around or below 50% of the heating power. It was dominated by low-Z impurities, whereas the W content was quite small. It is foreseen to operate ASDEX Upgrade several weeks without boronisation, which will allow investigating specific questions of D retention, radiation cooling and the influence of an uncoated W wall on the overall plasma

operation. Moreover, a large part of the whole experimental campaign lasting until August 2007 will be devoted to specific investigations concerning the W PFCs. Erosion and deposition of C in Pilot – G. Van Rooij

The linear plasma generator Pilot-PSI produces hydrogen plasma in a strong magnetic field (up to 1.6 T) to study ITER relevant PWI issues. Thomson scattering was employed at 17 mm in front of the target and yielded 1·1019 < ne < 2·1021 m-3 and 0.1 < Te < 4 eV. This demonstrates the ITER relevance of the experiment.

Carbon was exposed to the plasma of Pilot-PSI at different flux densities. Power measurements by calorimetry on the cooling water of the actively cooled target were shown to be in agreement with fluxes calculated from the Thomson scattering profiles in these measurements. This proved flux densities of 1-5·1024 m-2s-1 at a plasma temperature of 1.2 eV. Profilometry of the exposed targets showed an eroded crater and a zone of redeposition at the edge of the crater. The volume ratio between eroded and redeposited material was 2:1. The power to the target set the surface temperature between 600 and 1500 K. Relative measurements of the chemical erosion by emission spectroscopy as a function of target temperature yielded maximum erosion at the lowest temperature and a factor of 10 lower yield for T>1200 K. QMB studies of material migration in the JET sub-divertor region – A. Kreter In the recent JET campaign, the quartz microbalance (QMB) technique was applied to investigate the transport of carbon to remote areas in the divertor during disruptions. In each of “unwanted” disruptions recorded, a substantial additional amount of carbon was deposited in both inner and outer louvers. This effect seems to be not sensitive to the B field configuration, in contrast to the “normal” plasma operation with a predominant line-of-sight transport from the strike point position. It indicates a broadening of the energy deposition pattern during disruptions. The transported amount of carbon increases strongly with increasing plasma energy just before the disruptions. If we assume, that every JET pulse is terminated by a medium-size disruption, the amount of long-term stored carbon increases by a factor of ~3. Application of optical technique for in situ surface analysis of carbon tiles – T. Tanabe Optical absorption/emission spectroscopy, as a possible candidate method for in-situ characterization of plasma facing carbon tiles and quantitative evaluation of tritium, is applied to hydrogen implanted graphite and redeposited carbon layers in tokamaks. Because of semi-conductive nature of carbon, penetration depth of visible and infra-red light is so thin to make bulk analysis difficult. Nevertheless, optical absorption in visible wave length range, FT-IT and laser Raman spectroscopies give lots of information for surface layers of with in a few mm depth.

Appendix B: Participants N. Asakura N. Ashikawa X. Bonnin S. Brezinsek A. Chankin D. Coster G. Counsell T. Eich G. Federici M. Fenstermacher W. Fundamenski C. Grisolia A. Grosman M. Groth A. Herrmann J. Hu A. Kallenbach H. Kawashima V. Kotov A. Kreter K. Krieger A. Kukushkin V. Kurnaev K. Lackner B. Lipschultz S. Lisgo A. Loarte C. Lowry R. Maingi G. Matthews M. Mayer T. Nakano O. Ogorodnikova H. Pacher R. Pitts V. Philipps L. Pranevicius D. Reiter V. Rohde J. Roth M. Rubel M. Shimada C. Skinner (remote) P. Stangeby K. Sugiyama F. Subba T. Tanabe E. Tsitrone G. Van Rooij D. Whyte (remote) M. Wischmeier G. Wright R. Zagorski