**6th ITPA meeting on SOL/divertor physics**

The meeting was held over the period July 4-7, 2005 at the Facultat de
Ciències Jurídiques de la Universitat Rovira i Virgili in Tarragona. The
local coordinator was Carlos Hidalgo. The meeting lasted 3-1/2 days and
concentrated primarily on materials issues (see agenda in Appendix A).
There were \~ 42 participants (Appendix B).

There is an executive summary (section 1) followed by a more detailed
summary of the talks themselves (section II).

#### I Executive Summary

**D/T inventories and methods for D/T removal from surfaces:**

At the time of the 2004 Naka meeting it was recognized that fuel
retention in the gaps between tiles is significant (up to 50% of total
retention). At this meeting the results of IEA/ITPA collaborative work
investigating this further was reported. The deposition depth down the
gap varied with gap type (toroidal vs poloidal). The amount of
deposition appeared to correlate with fluence to the front surface.
These observations along with the driving physical process, assumed to
be hydrocarbon neutrals) will be tested in future work.

D/T removal was an emphasis of this session. Four types of oxygen
cleaning have been investigated with glow discharge cleaning appearing
to be the most efficient. However, ICRF was almost as good and could be
applied with the magnetic field on in ITER. More work is needed on the
recovery of good plasmas (time taken and the level of O in surfaces and
plasmas) following the cleaning, the efficacy at lower temperatures, the
applicability to mixed material layers, whether it properly cleans down
tile gaps, and whether the removal rate can be increased to that needed
for ITER.

The removal rates for photo-cleaning are approaching that needed for
ITER. When the deposits are thick high deposited energies are needed
which raises worries about dust creation and how much T would be
retained in the dust. There are three potential methods -- flashlamps,
lasers, and disruptive discharges. Besides the question of dust the
group is planning to look into questions of power requirements, the
applicability to mixed material surfaces and compatibility to magnetic
fields.

**Dust:**

Dust is a concern for ITER from the point of view of safety. A secondary
question is how the dust affects the core plasma in terms of impurity
level. The amount of work in this area has rapidly increased over the
last year. A range of presentations covered the formation of dust (e.g.
agglomeration in the plasma, diffusion-limited aggregation on surfaces,
and formation of quasi-homogeneous films). Modelling shows acceleration
of dust particles to very high speeds (up to a km/s) which allows them
to penetrate through the separatrix. Novel diagnostic techniques were
shown which allow measurements of the dust falling on surfaces and in
the plasma. The level of dust needed to halt operation was found to be
substantially higher than normally found in current tokamaks. More work
is needed to better understand the distribution of dust sizes, how the
use of mixed materials affects their behavior, and how the dust affects
the core plasma.

**High-Z experience:**

The effects of a variety of materials is being addressed at different
levels within the world program; from JET, with the initial mix of
materials, to ASDEX-Upgrade with steadily increasing amounts of W mixed
with C (and B), to C-Mod (fully high-Z, with B coatings). In all cases
we need a better understanding of D retention for individual materials,
the effect of mixed materials on retention/removal, and the effect on
core plasma operation (impurities and confinement).

All current tokamaks utilize wall-conditioning (e.g. boronization, Be
gettering) to achieve the best energy confinement regimes. ITER
performance is based on these results while at the same time it is
unlikely that ITER will have any wall conditioning similar to current
machines. Lastly, since it is currently thought that any reactor will
have all high-Z PFCs to reduce neutron damage and radiation levels, it
is considered important by the committee that at some point ITER convert
to all high-Z PFCs. To address the importance of conditioning C-Mod
removed all B from the machine (full coverage with Mo PFCs).
Unfortunately, operation without B was much poorer in terms of energy
confinement and core high-Z levels. Positive results were reported for
minimal adsorption of water after a vacuum break and general D
retention.

ASDEX-Upgrade results show clearly that the intermediate case of W, C,
and boronized surfaces can work well; low W level H-mode discharges can
be achieved with boronization and central heating to keep impurity
accumulation down. However, when there is no boronization, as after the
initial installation of W surfaces in the main chamber, core plasma
performance was degraded. ASDEX-Upgrade plans to finish the conversion
to all W PFCs over the next several years.

He bubble and blister formation in W has been shown to be dangerous for
W surfaces in plasma simulators. Tokamak experimental results do not
reflect this. Further work is needed to clarify the importance of this
process.

**First wall loadings (joint meeting with MHD group):**

During the last several years the committee has worked towards a better
understanding of first-wall particle and heat loadings. Much progress
has been made in identifying the various channels: a) steady state,
fluctuation drive fluxes; b) ELM fluxes; & c) disruptions. Based on a
review of disruption power loadings it was apparent that the original
assumptions used by the ITER group are incorrect in a number of points;
e.g., the fraction of core plasma stored energy arriving at first-wall
surfaces, the time scales.

The understanding of first-wall loadings during ELMs has advanced
considerably in the past year. The filamentary structure
(toroidal/poloidal number, rotation velocities) as well as the
transmission through the SOL have been characterized. New measurements
of the localized heat loads on first-wall structures have begun to
emerge. The ELM energy that goes to the wall is localized toroidally and
poloidally leading to localized erosion/melting of limiter structures.
The localized nature of the power loading makes the characterization of
heat loads (frequency, total heat load) difficult.

Disruption mitigation is being demonstrated to be an important tool for
extending divertor lifetime. It was shown that gas could be successfully
used to penetrate plasmas with pressures similar to ITER. New
measurements (imaging, MHD, Thomson scattering, bolometry) are providing
much-needed data for comparison with codes being applied (MHD and plasma
evolution).

Given the importance of this session's subject it was agreed that
position papers summarizing recent progress would be written for the
ITER group on disruption power loads, ELM power loads and disruption
mitigation strategies.

**Mixed materials**

Utilizing different materials for different surfaces has the advantage
of being able to tailor their capabilities for different applications.
Recently it has been noted that there are unintended consequences as
well; both positive and negative. Be deposited on C tends to reduce C
chemical erosion. Be layers will retain less T than C layers and the T
can be removed at lower temperatures. However, Be layers on W have
negative consequences. Thin layers of Be~12~W form which, if they
remain, have low melting temperatures and poor thermal contact to the
bulk W below. What is not clear is whether such layers will form in
regions where the deterioration of surface characteristics will reduce
divertor lifetime. In other words it is possible that in high heat flux
areas the layers will not form because Be will be re-eroded before it
can diffuse into the W.

**Be operation in ITER**

This session was aimed at taking current knowledge of first-wall heat
loads, material erosion and migration, and predicting the performance of
the Be first wall in ITER. The goal was to examine whether the community
could reach agreement of how Be would affect ITER operation. We also
hoped to examine the basic assumptions going into these estimates and
determine if more work could be done to verify them. Both extrapolations
of current experimental results and modelling were presented.

The current estimates of first-wall erosion cover a wide range from low
to 1 μm/discharge, thus placing some limit on first-wall lifetime and
being a significant impurity source.

The general expectation is that if ELM heat loads on the divertor can be
reduced to acceptable levels, the same may be true for ELM erosion of
the first-wall surfaces. The same cannot be said about disruptions
where, even if the energy is spread over a large fraction of the
first-wall during the thermal quench (mitigated or not) there is likely
to be significant melting.

It was generally agreed that the majority of the eroded Be goes to the
inner divertor region. This is of course based on current C and Be
migration measurements. However, the details of how it reaches that
location and whether the differences in ITER divertor will affect the
prediction are certainly highly uncertain.

The replacement of C with Be as the main eroded material is likely to
reduce T retention from that predicted for a full C ITER. The
concentration of T in Be is typically in the 1-5% level (can be 40-100%
in C). T is released from Be at lower temperatures (600 oK) than for C.
Finally, once Be migrates to the inner divertor, it does not appear to
migrate to sheltered areas that are difficult to access with T cleaning
techniques. That said, it is not clear how well T can be removed from
mixed T/W/C layers and whether Be-W alloys will lead to a degradation of
the W divertor capability in high heat flux areas.

Once first-wall Be surfaces are damaged (by ELMs of disruptions) we are
concerned about how much the heat-handling capability is reduced. This
is a particular concern for the upper divertor and the startup limiter.
Be (and W) melt-layer dynamics in real tokamak situations need to be
better understood as well as the resiliency of the core plasma to absorb
large amounts of low-Z ions for short periods.

**Discussion of the ITER dome**

No consensus was reached on whether removal of the dome would enhance or
hurt ITER operation. There exists no experimental evidence or modelling
results indicating that the dome reduces neutral backflow to the plasma
(part of the original motivation). At first glance this result together
with concerns about dust and T collecting under the dome would seem to
argue for its removal. However, a detailed engineering design for the
dome replacement does not exist. It is thus difficult to assess whether
dust removal and shadowed areas would be significantly reduced. Finally,
the removal of the dome has very negative implications for diagnostics
that will be compromised or removed. If a replacement for the dome is
designed and can retain diagnostics the SOL/divertor committee stands
willing to help evaluate its performance.

Next ITPA meeting and topics

It was agreed thatsince the meeting had not been held in Asia for at
least a year we should have one there. The most convenient time is to
combine it with the Plasma Surface Interaction meeting abstract
selection meeting which already includes a number of our members. If
Jiangang Li agrees to this it will likely occur either in China or Japan
in late January or early February. Now that the US is being allowed to
host meetings again it is possible that the following meeting would be
held somewhere in the US in the Summer of 2006.

A number of topics were suggested for the next meeting. They include D/T
retention in surfaces, wall conditioning, startup studies, flow
characterization, melt layer stability, vapor shielding, and
spectroscopy of hydrocarbons in low-T plasmas.

We are preparing position papers advising ITER or revised assumptions
for first-wall and divertor loadings under ELM and disruption
conditions. We are also planning to revise the high-priority task list
to reflect an ITPA CC request for more near-term goals.

### 

### II. Detailed session summaries

### Session 1: 'D/T inventories (surfaces and sides of tiles) and their removal

### 1.2 'Recent results on carbon deposition and fuel retention in gaps of plasma facing structures', K. Krieger

Plasma facing components for the ITER divertor and baffle regions will
likely be manufactured with macrobrush structures or castellated
surfaces. Material samples with gaps of similar geometry were exposed to
different plasma conditions in ASDEX Upgrade, TEXTOR and DIII-D. In all
devices an exponential decrease of both carbon and deuterium inventories
at the side faces from the gap entry into the gap is found. The scale
length is mainly related to the gap width and the orientation of the
B-field. The fraction D^+^ incident on the front surface which is
retained in the gaps is in the range of a few percent. Extrapolation of
growth rates under low flux conditions to ITER dimensions shows no
significant contribution to the total tritium inventory. Growth rates
observed in gaps exposed to high fluxes such as near the strike point
zones are 100 times higher, which might be of concern for ITER operation
with carbon PFCs. On the other hand, experiments with samples kept at
elevated temperature (200ºC) show a decrease of retained D by a factor
of up to 10, although this observation is restricted to detached plasma
conditions.

1.3 'Boron & Deuterium down tile gaps in C-Mod', D. Whyte

Ion beam analysis was used to measured boron and deuterium deposition
down tile gaps of the C-Mod Mo tiles after \~7 years of exposure. Boron
is applied frequently to Mo tiles using ECDC of He + diborane with a
weak magnetic field (\~900 Gauss) Boron layers up to \~1 micron thick
were found on edges near the plasma-facing side with typical e-folding
distances of 1-3 mm down the gaps, which is 2-6x the distance between
tiles (0.5 mm). Deuterium deposition clearly correlated to B, indicating
codeposition with B as the main reason for D retention. The D/B ratio
varied from \~2-10% with an average value of 5% in most cases. The D/B
ratio is similar to those found on the plasma-facing tiles. It is
estimate that 10-20% of the deuterium retention was found in the gaps
for the C-Mod tiles. Toroidal running gaps were found to have deeper
penetration of boron and deuterium down the gaps than in the poloidal
running gaps. Variation was also found in e-folding distance as a
function of poloidal location of the tiles. This suggests that B
deposition, along with D codeposition, is occurring during plasma
discharges.

1.4 - 'Recent results on carbon migration and deposition in ASDEX
Upgrade', V. Rohde

During the 2002/03 campaign a divertor marker experiment to get a
complete data set on the carbon migration and deposition was performed
in ASDEX Upgrade. Even below the divertor structure deposition and
erosion are observed. As the erosion is direction sensitive, it is
assumed that ions play an important role. Parasitic plasma observed
below the roof baffle can produce these ions. The surface loss
probability at remote areas is determined by cavity probes: gamma =
0.65-0.75. This value is consistent with the observation that deposition
on the qmb\'s is strongest if the strike point position is in the line
of sight of the qmb location. The temperature dependence of the
deposition was investigated by heated and cooled probes. As expected for
re-erosion by atomic hydrogen the layer thickness can by reduced by a
factor of 50 for a probe temperature of 200 C. Probes mounted at the LN2
shield of the in vessel cryo pump measure a factor of 100 enhanced
deposition. The deposition on the target plates shows significant
differences for short term (^13^CH~4~ puffing) and long term deposition.
A peak at the inner baffle, caused by direct ionisation of the injected
^13^CH~4~ is not observed on long term data. Whereas the ^13^CH~4~ data
shows a stronger deposition at the outer divertor, the long term data
shows the dominant deposition at the inner divertor. Obviously
redistribution of the deposited carbon has to be taken into account to
understand the long term deposition pattern. QMB data show constant
layer growth at the inner divertor structure. At the outer divertor
erosion and deposition phases are observed.

1.5 - '^13^C injection at the outer divertor in JET', V. Philipps

^13^CH~4~ injection into the outer SOL of JET under identical H-mode
discharges show about 10% of injected ^13^C deposited on the inner
tiles, preferentially on the horizontal tile, while the strike point was
on the vertical tiles. The transport towards the inner divertor seems to
originate from ^13^C that has entered the confined plasma , diffused to
the SOL and driven by SOL flows to the inner divertor. Interestingly the
^13^C has undergone further transport in the inner divertor from the
vertical strike point tile to the horizontal tile which was in the PFR
which is probably due to Elms. The amount found on the vertical tiles
near the outer injection is comparably small (about 15%), but 2
horizontal outer divertor tiles have not been analysed so far.

1.6 - 'Recent activities on photocleaning in the EU', C. Grisolia

In the EU, photon detritiation using flash lamp or laser ablation
techniques is under strong development. Flash lamp detritiation has been
tested in JET and films have been removed with a rate of 2.5 to 10 m^2^
of 10µm co-deposited films per hour with the flash lamp in the ablation
regime at 250 Joules per pulse. Laser detritiation technique has been
proven at a laboratory scale with a rate of 1 m^2^ of 50µm deposited per
hour using a 100 W laser. However, more R&D is necessary to meet the
constraints of ITER.

1.7 - \"Oxygen cleaning activities in TEXTOR\", A. Kreter

Oxygen cleaning techniques were applied in TEXTOR to remove co-deposited
carbon/hydrogen layers. The following table gives an overview of
achieved C removal rates, main advantages and drawbacks of the
techniques as well as possibilities to increase the C removal rates. C
removal rates have to be compared with an integral TEXTOR carbon
redeposition rate of 2.7·10^20^ C/s

+-----------+-------------+----------------+--------------+-----------+
| Technique | C removal   | Advantages     | Drawbacks    | Possible  |
|           | rate        |                |              | imp       |
|           |             |                |              | rovements |
+-----------+-------------+----------------+--------------+-----------+
| Oxygen    | 2.5·10^18^  | > Simplicity   | > Low        | > Higher  |
| venting   | C/s for 0.3 | >              | > removal    | > O       |
|           | mbar,       | > Access to    | > rates      | >         |
|           | Twall=620 K | > all wall     | >            |  pressure |
|           |             | > areas        | > T~wall~ \> |           |
|           |             | >              | > 600 K      |           |
|           |             | > Selective    | > needed     |           |
|           |             | > removal of   |              |           |
|           |             | > redeposited  |              |           |
|           |             | > layers       |              |           |
+-----------+-------------+----------------+--------------+-----------+
| Glow      | 2-3·10^19^  | > Higher       | Incompatible | > Higher  |
| discharge | C/s         | > removal      | with steady  | > GD      |
| con       |             | > rates        | state B      | > current |
| ditioning |             | >              | n            | >         |
|           |             | > Applicable   | on-selective | > Higher  |
|           |             | > for low      | carbon       | > pumping |
|           |             | > T~wall~      | removal      | > rate    |
|           |             |                |              |           |
|           |             |                | > Limited    |           |
|           |             |                | > wall area  |           |
|           |             |                | > access     |           |
+-----------+-------------+----------------+--------------+-----------+
| ICRF      | 1.8·10^19^  | > Higher       | > O          | > higher  |
| con       | C/s for     | > removal      | > injection  | > pumping |
| ditioning | 1:10 duty   | > rates        | > limited by | > rate    |
|           | cycle for   | >              | > pressure   | > (also   |
|           | pump out    | > Applicable   | > limit at   | > for     |
|           |             | > for low      | > antenna    | > steady  |
|           |             | > T~wall~      | > box        | > state   |
|           |             | >              | >            | > ICRF)   |
|           |             | > Compatible   | > N          |           |
|           |             | > with steady  | on-selective |           |
|           |             | > state        | > carbon     |           |
|           |             | > magnetic     | > removal    |           |
|           |             | > field        | >            |           |
|           |             |                | > Limited    |           |
|           |             |                | > wall area  |           |
|           |             |                | > access     |           |
+-----------+-------------+----------------+--------------+-----------+

1.8 - 'Oxidation experiment in HT-7', J.S Hu

The oxidation experiments, including O-ICR conditioning, O-GDC and
O-baking have been done in the HT-7 superconducting tokamak. The
temperature of the limiter is 402-425K and that of the liners is
435-470K. The higher pressure and conditioning power are favorable for
removal of hydrogen and carbon. Pure oxygen and gas mix of He/O were
used. The highest removal rates of H, D and C-atoms in pure O-ICR
experiment up to 2.64x10^22^, 7.76x10^21^ and 1.49x10^22^ atoms/hour
respectively were obtained in 40kW 9x10^-2^Pa O-ICR cleaning,
corresponding to the removal rate of co-deposits of about 317nm/day
(7.2g/day for carbon). The highest removal rates of H and C atoms in gas
mix He/O-ICR experiment up to 5.4x10^21^ and 7.2x10^21^ atoms/hour,
respectively, were obtained in 4:1 40kW 9x10^-2^Pa He/O-ICR cleanings.
Average removal rates, 5.2×10^22^ H-atoms/hour, 5.65x10^21^ D-atoms/hour
and 5.53x10^22^ C-atoms/hour, respectively, were obtained in 145min
O-GDC experiment in a pressure range of 0.5\~1.5 Pa.. Only O-ventilation
has very little effect to remove co-deposited layers and hydrogen. By
correct cleaning, most oxygen retained on the wall was sufficiently
removed before plasma discharge. Plasma discharges could be recovered
after a few tens of shots with large disruptions. Even though the
impurities, such as C, O increased, the hydrogen recycling was largely
reduced by He/O~2~-ICR experiment. No apparent damage in the torus after
oxidation experiment was observed.

1.9 - 'ICRF H/D removal in LHD', N. Ashikawa,

ICRF wall conditioning under high magnetic fields has been started in
LHD. Operation was achieved with P~icrf~ from 8 to 149 kW and helium
pressure from 10^-2^ to 10^-1^ Pa.

In these initial experiments sufficient power was absorbed by hydrogen
as determined from electron density and FNA data. An increase in
hydrogen pressure with RF power was observed. LHD did not have RF
breakdown in the range 10^-2^ to 10^-1^ Pa in He pressure. The hydrogen
removal rate by GDC is larger than with ICC. Optimization of the ICC
with RF phasing is an important part of future work.

1.10 - 'Summary of nitrogen injection experiments in AUG', P. Tabares

Nitrogen injection experiments at the subdivertor region of Asdex
Upgrade have shown for the first time that a decrease of up to a factor
of 5 can be achieved in the carbon deposition rate at positions not
accessible directly by the plasma. Furthermore, the required nitrogen
flows are fully compatible with the unperturbed operation in H mode
discharges. No change in the deduced carbon concentration in the
divertor plasma has been detected so far, in line with the expected
scavenger effect of the injected species.

1.11 - 'Effects of wall saturation on plasma operation', E. Tsitrone

The main effects of saturation of carbon walls on plasma operation are
: 1) loss of the density control during a discharge; 2) difficulties to
start up the plasma (no sustained breakdown).

The first effect has been clearly observed in all tokamaks going to long
pulse operation with a non or partly actively cooled machine (Tore Supra
before the CIEL upgrade, JT60U, HT7 ...), resulting in an uncontrolled
density rise (typically \< 1 minute); temperatures of the plasma facing
components rise during the entire discharge.

The CIEL upgrade to a completely actively cooled machine dramatically
improved the density control in Tore Supra, allowing discharges up to 6
minutes/3 MW/1 GJ and 1 minute/7 MW/0.4 GJ. Conditioning is needed
before reaching reproducible conditions for the high power/high density
shots. This is attributed to plasma facing components not in direct
interaction with the plasma, with long thermal time constants.

Wall saturation experiments running repetitive discharges have been
performed on JT60U and Tore Supra. The wall pumping capacity is seen to
decrease from shot to shot, and the series of shots end up with a
disruption, and subsequent not sustained breakdowns.

To overcome this problem, techniques to monitor the wall saturation
status have been successfully developed in Tore Supra. Cleaning
techniques to restore the wall pumping capacity use cycles of very low
current (\< 40 kA) successive breakdowns. This process is compatible
with the toroidal field, but is not directly applicable to ITER as the
magnetic coils system is different.

In ITER, which will run under very high D fluence within one discharge,
the consequences of wall saturation on plasma operation should be better
assessed. Operation of most present day machines relies heavily on
conditioning techniques (boronisation, Be evaporation, He glow
discharges ...), and that the best performances in terms of confinement,
on which the extrapolation for ITER is based, have generally been
obtained with pumping walls. ITER relevant techniques to monitor the
wall saturation status and optimise the start up should therefore be
developed.

1.12 - 'Summary of the Session 1 discussion', V. Philipps

**Material and fuel retention in gaps:** all the present data show that
material and fuel deposition in gaps of ITER wall components will
seriously contribute to the tritium retention. In order to extrapolate
more quantitatively from present data to ITER, the mechanism of how the
material and fuel penetrate into the gaps must be better clarified. An
important question is whether this migration is mainly restricted to
carbon, due to the special carbon chemistry, or is also expected e.g for
Be. Presently, it is believed that the hydrocarbon chemistry is
important for the gap deposition, but more studies are necessary. The
other important question is how to extrapolate from present data to
ITER. The fraction of fuel deposited on top of the tile surfaces
compared to that in gaps is sometimes used. A perhaps more appropriate
quantity to use is the fraction of material and/or plasma fuel arriving
on the surface compared to that trapped in gaps. Some data presented
showed 0.2 % of the surface D fluence being retained in the gaps of an
ITER-like wall component (TEXTOR) but the field line inclination was
steeper (20°) than that expected in ITER More experiments elucidating
the transport mechanism are necessary.

**Material migration**: while the overall tendency for the inner
divertor to be a deposition dominated area is common to all divertor
tokamaks, the behaviour of the outer divertor is still difficult to
predict. DIID and ASDEX-Upgrade data show that erosion and deposition
dominated phases are both observed at the outer divertor. Recent data
show consistently in DIID and AUG a strong influence of the substrate
temperature on deposition rates, with a decrease by about one order of
magnitude from ambient to 200°C. This calls for hot surfaces at the
deposition area in ITER, mainly the dome and PFR region. The transport
of material from the outer to the inner divertor is still an open issue
and needs further clarification. In particular, particle flow through
the PFR may play a role. Analysis of recent 13C injection into the outer
divertor region of JET is not yet available.

**Fuel removal**: development of fuel removal methods that are
applicable in ITER environment is most important. Photo-cleaning has
been demonstrated on larger areas in JET but concerns remain about the
concurrent production of dust (both for T retention and for safety
issues of large quantities of C dust). Dust collection in parallel with
photo-cleaning is absolutely mandatory but may be difficult and not 100%
efficient. Oxygen cleaning can only be used for carbon layers. It is
assumed presently that GDC and ICRH will clean only the plasma facing
sides. The removal action of these standard methods in hidden areas
should be investigated. Oxygen molecule oxidation of C layers needs
temperatures above 250-350°C, as shown again in the HT-7 tokamak. Such
temperatures are extremely difficult for ITER to achieve. Plasma
recovery after oxygen treatment needs to be documented and scaled to
ITER. Scavenger techniques like the use of N may be of use but are far
away from being a proven technique for ITER.

**Density control and wall saturation** becomes more and more important
with increasing pulse length as seen in Tore Supra and other machines.
However, based on our present understanding, the importance of such
effects for ITER which has a larger particle throughput (needed for He
exhaust) is not clear. More experiments are needed to develop a proper
understanding of the processes involved, how to control them, and how to
prepare the walls for such long pulses.

**Session 2: 'Dust'**

2.2 'Results of Dust in JT-60U & LHD' , N. Ashikawa

Based on dust measurements we have extrapolated the amount of dust in
the whole vessel as 7 g in JT-60U and 3-10g in LHD. The production rates
normalized to discharge time are 0.2 mg/s in both devices. The total
amount and production rate are similar in both devices. But the particle
size distribution was different. In JT-60U, dust deposition was found in
remote areas of the outer divertor region. Small dust with nm order size
can not be bypassed to the total amount by large dust with glowing
process. Moving particles were observed by high speed IR camera and this
transport time in the camera FOV is about 40ms and estimated
transporting speeds are about 20 m/s at the outside edge.

2.3 Development of dust diagnostics', R. Maingi, C. Skinner: ITER safety
will depend on knowing and controlling the inventory of dust on interior
surfaces, but diagnostics to estimate the dust inventory need to be
developed. A novel electrostatic dust detector has been demonstrated in
both air and vacuum environments in the laboratory \[Rev. Sci. Instrum.,
75 (2004) 370\]. A fine grid of interlocking traces with spacing down to
25 µm is biased with 30-50 v DC. Impinging dust produces a short circuit
and the resulting current pulse both vaporises the dust and produces a
pulse that is recorded with standard nuclear counting electronics. 
Particle size information can be obtained from the electrical waveform.
A detector has been installed at the NSTX at a port under divertor. Such
detectors could be adapted for specific areas in ITER. (Maingi for
Skinner et al.)

2.5: \"Modelling of Dust Formation in Plasmas\" - Xavier Bonnin.

A new edge plasma physics group has been formed at Université Paris
XIII, to work on issues at the confluence of industrial cold plasmas and
edge fusion plasmas. One such common research area is dust. We have
built a model for dust formation and nucleation, via carbon cluster
chemistry, and compared it to a DC Argon discharge experiment in
Marseilles where Ar sputters a Carbon cathode and dust is collected at
the anode. The dust is cauliflower-like and reaches cluster sizes of
order 40 nm. The model can reproduce such a formation by agglomeration
of C2 and C3 sputtered clusters onto larger ones and its experimental
time evolution to within a factor of two. More accurate chemistry rates
(in particular for negatively charged clusters) is expected to improve
the match to experiment. Additionally, a new plasma Chemical Ablation,
Sputtering, Ionization, Multi-wall Interaction and Redeposition
(CASIMIR) device is currently being built to address issues related to
hydrocarbon erosion products chemistry, transport and redeposition in
parasitic plasma environments as those expected under the ITER divertor
dome.

2.6: 'Dust in DIII-D', M. Fentermacher, D. Rudakov, P. West

Intrinsic dust in DIII-D has been characterized using signals from the
Rayleigh scattering channel of the core and divertor Thomson scattering
systems. A small particle (about 60 nm) distribution and a large
particle distribution (\> 300 nm) are indicated by the data. Average
dust particle density ranges from 0.005 particles/cc in the upper part
of the vessel to 10 times that level in the lower part of the vessel.
Dust injection experiments have also been done using 1-10m particles
on a DiMES sample. Tracks of dust particle motion when the dust
encounters the OSP show a preferential direction that is about 15-45
degrees inboard of the magnetic field direction. Drag force caused by
plasma flows, ExB drift in the pre-sheath electric field can explain
this direction. Some curvature is also observed in the particle
trajectories. These measurements provide a test dataset to validate dust
particle transport codes.

2.7: 'Operational recovery from enormous amounts of titanium dust', B.
Lipschultz

Because of an unfortunate degradation of the font end of the new C-Mod
LH coupler (made of Ti) a very large amount of Ti dust was created in
C-Mod., After removal of the waveguide \~ 300 g remained in the vessel
affecting startup and operation. This is equivalent to something in the
range of 24 kg (scaling by area) to 220 kg (scaling by volume) in an
ITER-size device. The length of plasma discharges increased and reached
full length over \~ 200 shots. The Ti level and the gas retained by the
dust dropped over a similar amount of time. Given the comparison to Mo
levels in the plasma it appears that normal amounts of dust are not
directly affecting the discharge impurity levels. Only when the dust
levels get very large is it a problem.

2.8 - 'Dust dynamics in MAST', GF Counsell (UKAEA Fusion, EU)

A model has been developed in association with Imperial College, London
which allows the motion and thermal properties of dust to be evaluated
in the fusion plasma environment. The code takes 2D data of plasma
properties (temperatures, densities, plasma flows etc) and the magnetic
equilibrium from standard output of the B2SOLPS5.0 fluid code. Monte
Carlo test particles can then the launched with an arbitrary mass and
velocity distribution from any poloidal location and followed until they
either evaporate in the plasma or strike a wall. Both tungsten and
carbon dust has so far been modelled. The dust charge at each time step
is calculated from OML theory, allowing for the impact of secondary
electron emission, using the local plasma parameters and the
acceleration is then evaluated using an equation of motion including the
Lorentz, pressure gradient, flow pressure and gravitational forces.
Temperature evolution of the dust is evaluated taking into account all
significant heating and cooling mechanisms (e.g ion, electron and
neutral bombardment, radiative and ablative cooling etc.). For a typical
dust trajectory, flow pressure is the dominant force when the grain is
large, with the Lorentz force dominating as the dust moves into regions
where evaporation occurs. For MAST plasmas, dust launched from a
toroidally and poloidally localised source is widely redistributed
around the vessel by the plasma and some dust trajectories can lead to
evaporation occurring inside the separatrix. Modelling of dust transport
in ITER plasmas is on-going, with the eventual aim of predicting regions
where dust is likely to accumulate in the vessel.

2.9 - 'Summary of discussion' - S. Krasheninnikov

**In situ experiments with dust:** Intrinsic dust in DIII-D has been
characterized using signals from Thomson scattering systems. Average
dust particle density ranges from 0.005 particles/cc in the upper part
of the vessel to 10 times that level in the lower part of the vessel.
Dust injection experiments have also been done using 1-10 micron
particles on a DiMES sample. Tracks of dust particle motion show a
preferential direction that is about 15-45% inboard of the magnetic
field direction and the estimates of dust particle speed give
\~10-100m/s. A similar magnitude for dust speed was also observed in
JT-60. In C-Mod a huge generation of dust due to Ti structures of
antenna resulted in a very short, resistive plasmas with no density
control etc. After removing Ti structures and boronization, and
following a few hundred shots, standard core performance was recovered.
\~ 300g of dust still remains in vessel which  may indicate that the
remaining dust is localized in the regions with rather small exposure by
plasma. Since the Mo dust level is much lower and Mo is affecting the
core plasma more, the inference is that dust is normally not a
contributor to steady state impurity levels. Dust is likely to be more
important for 'injections'.

**Analysis of dust collected in fusion devices:** Dust production rate
\~ 0.2 g/s was found in both JT-60 and LHD. Taking into account the
erosion rate of divertor tiles in JT-60 outer divertor it gives \~ 7%
transformation of eroded carbon into dust;

**Development of dust diagnostics:** A biased fine grid with the spacing
down to 25 microns biased to \~50 V has been developed for measurement
of dust landing on it. It works fine in a test chamber. Particle size
information can be obtained from the electrical wavefront. The detector
is installed in NSTX at a port under the divertor. An SEM measurement,
allowing in situ monitoring of the first wall structure, will be
installed at JET;

**Dust modeling:** To model the dynamics of dust in tokamaks 3D codes
(DUSTT at UCSD and another one at Culham/Imperial Collage) are being
developed. Codes allow tracking of a test dust particle in toroidal
geometry with realistic plasma background. Preliminary results of dust
transport modeling in NSTX, DIII-D, ITER and MAST tokamaks have been
reported showing reasonable agreement with available experimental data.
It was shown that dust obviously penetrates deeper into edge plasma than
individual neutral impurity atom causing a significantly bigger impact
on both edge plasma parameters and detachment processes. A new edge
plasma physics group, at Université Paris XIII is to build a model for
dust formation and nucleation, via carbon cluster chemistry, and compare
to experiments in Marseilles. A new plasma device CASIMIR is currently
being built to address issues related to the chemistry, transport, and
redeposition in parasitic plasma environments such as those expected
under the ITER divertor dome.

# 

**Session 3: 'High-Z experience'**

# 3.1 - 'W operation in ASDEX Upgrade', A. Kallenbach -- IPP Garching

ASDEX Upgrade is currently coated with 70 % tungsten PFC coverage, it is
planned to come to a complete coverage within the next 2 years. The
dominant sputtering source is caused by low-Z impurity ions of C, B and
O. Strong variations of the peripheral W concentration (measured around
Te= 1 keV) and the central W concentration are observed. Divertor
operation is affected under 2 experimental conditions due to strong
central W radiation: in the low power H-mode with low ELM frequency,
impurities are driven from the SOL to the pedestal by a strong inward
drift. The resulting strong radiative losses may cause a H-L transition.
This behaviour can be cured by pellet ELM pace-making. For discharge
conditions with good confinement and central electron density peaking
inside ~p~= 0.4, strong central accumulation of W occurs. This is
driven by neoclassical effects and can lead to central radiation
approaching the heating power density. Central heating above a power
threshold flattens the central fuel density profile and consequently
leads to flat tungsten profiles with W concentrations around 10^-5^.

3.2 - 'High temperature erosion and melting of tungsten in TEXTOR', A.
Kreter

Experiments on high temperature erosion and melting of tungsten were
performed in TEXTOR. Observed tungsten erosion was attributed to the
sputtering by carbon and to the sublimation. No indication of an
enhanced high temperature erosion of tungsten was found. A local
tungsten flux for T~surf~\<3000K of up to 4x10^21^ atoms/m^2^/s was
measured, corresponding to 4% of the background flux, in agreement with
previous observations in TEXTOR. No increase of the tungsten
concentration in the core was observed. Power deposition onto the
tungsten plate resulted in a melt layer formation and a fast melt layer
motion with a velocity of \~1.5 m/s. The motion of molten tungsten can
be attributed to the thermo-emission current and the resulting JxB
force. Surface studies indicated a recrystallization of tungsten with
the maximum grain sizes in the zone of the highest temperatures. Carbon
dust particles were found incorporated into the melting zone. No
blistering in the melting zone and other parts of the tungsten plate was
observed.

3.3 - 'Removal of boron from Alcator C-Mod and the effect of
boronization', B. Lipschultz

All currently operating tokamaks use some kind of PFC coating - whether
it be boronization (e.g. ASDEX-Upgrade, C-Mod, DIII-D, JT-60U.....).
ITER performance is based on this while at the same time ITER presently
will not be able to use such wall coatings. Information is needed on
what the wall coating provides. Unfortunately, coatings in current C PFC
machines cannot be readily removed to make this comparison. C-Mod has
provided unique results in this sense by removing the B coating and
examining changes in performance. In addition this provides data for an
all high-Z first-wall, under consideration for ITER. The results
indicate that the B coating is necessary to reduce the effect of Mo on
core performance. At the beginning of H-mode the impurity confinement is
essentially infinite and any Mo source is seen in the plasma, often
leading to an H-L transition. The B layers also play a role in adsorbed
D, lengthening the pumpdown. On the other hand the B layer does not
affect the absorbed D compared to a bare Mo surface. Ongoing experiments
are being made to investigate inter- and intra-shot boronization in
support of ITER. The implication for ITER is that all high-Z PFCs might
lead to significant core Mo concentrations during H-mode. Whether this
will have an adverse effect on confinement and the exact level of W
requires further study.

3.4 - 'W-tile experience from JT-60U', Y. Ueda/T. Nakano

Tungsten divertor tile experiments were started in JT-60U in the
experimental campaign 2003-2004. 13 tungsten coated CFC tiles were
installed in the P-8 section (12 tiles) and the P-17 section (1 tile)
just above the CFC tiles where the outer strike points normally
intersect. Plasma heat load up to 6 MW did not cause any notable damage
on W-tile surfaces. On the CFC tile toroidally adjacent to the W-tiles,
localized thick W deposition near the W tile edge was observed. Except
for this area, almost uniform thin W deposition was observed. W
migration to the surface of the dome top tile in the P-8 section was
observed, while no observable amount of W was found on the inner baffle
tiles. In high NBI power shots (\~15 MW), W accumulation could take
place. This accumulation could be mitigated by edge gas puffing
(preliminary results).

3.5 - 'He ion irradiation effects on W', Y. Ueda/N. Yoshida

Tungsten samples pre-irradiated by 8 keV He at RT to the fluence of up
to 10^22^ m^-2^ were exposed to 13 MW/m^2^ heat load by the electron
beam. It was found that He pre-irradiation strongly deteriorated thermal
shock resistance of W. In the case of 10^22^ m^-2^, erosion up to the
depth of 0.8 m (10 times larger than the He ion range) took place due
to reduction of thermal conductivity and hardening by He bubble and
blister formation. He bubbles were also formed by low energy He
irradiation under the conditions of T \> 1300 K, fluence \> (10^25^
\~10^26^) m^-2^, and He energy \> 6 eV. Since edge plasma parameters on
the ITER outer divertor will satisfy these conditions, He bubble
formation on the outer tungsten diverter plates will likely occur at
elevated temperatures and affect the operating conditions (surface
temperature limit, allowable ELM heat pulse energy, etc.). Intensive
works are needed to evaluate the impact of He irradiation to W for the
material selection of ITER divertor.

3.6 - Discussion and summary of Session 3 "High-Z experience", A.
Kallenbach

During the sessions talks, several new results were presented on tokamak
operation with high-Z plasma facing components: Operational restrictions
imposed by tungsten radiation have been observed in AUG with 70 %
tungsten coverage. These can be cured by central heating for density
profile flattening and ELM pace-making to avoid long ELM-free phases.
JT-60U observed preferential W migration from tiles installed in the
outer divertor in toroidal direction as well as to the dome. The
behaviour of W erosion under high heat fluxes was investigated in
TEXTOR. Tungsten erosion could be attributed to sputtering by carbon and
sublimation, and no anomalously high yields were observed at high
surface temperatures. These experiments were driven up to W melting, and
the motion of the molten W could be attributed to thermo-emission
currents and the resulting jxB force. No W blistering was found.
Tungsten samples exposed to high He fluences in NAGDIS-II showed bubble
and blister formation. During the discussion, different possibilities
were brought up why these negative effects occur in plasma simulators,
but not under tokamak conditions. One reason could be the fact that
monoenergetic He beams produce a different deposition characteristic in
the material. Further work is needed to clarify this topic. The
importance of wall conditioning in a full-metal device is a topic which
is highlighted by recent results from Alcator C-Mod, where degraded
plasma performance was observed during non-boronized operation with Mo
walls cleaned from B.

In the discussion session, it was emphasised that a strategy needs to be
developed for the possible implementation of additional high-Z PFCs in a
later ITER phase, which takes into account technical boundary conditions
like the possibility of a change of the first wall. While a much lower
fuel retention by tungsten is expected compared to carbon, the reduction
of divertor radiation will require additional impurity seeding. Two
topics are foreseen to be put on the agenda for the next meeting: 1)
possible impact of a high-Z startup limiter (erosion, flux consumption)
and 2) wall conditioning of an all-metal device.

**Session 4: 'First wall loadings', joint meeting with the MHD group**

4.1 'EU-PWI Task Force Studies on Disruption Energy Fluxes', A. Loarte

Analysis of the energy balance and timescales for energy fluxes during
similar disruptions in JET, ASDEX Upgrade, MAST, DIII-D, TEXTOR and FTU
was presented. On the basis of the experimental data analysed, the
following conclusions have been reached :

> a\) The thermal energy of the plasma at the thermal quench is
> significantly lower than that during the full performance phase of the
> discharge for most disruptions, typically by a factor of 2 to 4.
>
> b\) There is a favourable size scaling of the timescale for energy
> flux to PFCs during the thermal quench, but there is a significant
> spread in this timescale (by factors of \~ 6) within each experiment
> and a large variation disruption-to-disruption for nominally identical
> discharges.
>
> c\) There is a significant broadening of the heat flux during the
> thermal quench (by a factor of 5-10 compared to the full performance
> plasma) in diverted discharges. This seems to be absent in limiter
> disruptions.

These new measurements call for a re-evaluation of the conditions
expected for the average disruptions to be encountered when operating
ITER in its reference Q~DT~ = 10 scenario based on the ELMy H-mode. In
the first place, the average plasma energy at the thermal quench in ITER
will be \~ 88 MJ to 175 MJ, if no disruption amelioration is applied,
i.e., \~ 25 -50 % of the full performance plasma energy. This energy
will flow to the divertor over a very large area, which will be
typically in the range of 5-10 times the divertor wetted area for power
flux in steady-state conditions (A~s.s~), i.e. 18--35 m^2^ in ITER. The
typical timescale for the thermal quench power pulse will be in the
range 1.5--3 ms. As a consequence of these factors, the expected
divertor energy flux in the average ITER disruption, if amelioration
actions are undertaken by the systems, will be \~ 3.3 MJm^-2^, with a
timescale of \~ 2.3 ms. Using the standard deviations in the
distribution functions, the lowest and highest thermal quench energy
fluxes will be \~ 7.5 MJm^-2^ in \~ 1.5 ms and \~ 1.3 MJm^-2^ in \~ 3.0
ms, respectively. These values have to be about a factor of \~ 2 higher
if only the natural degradation of plasma confinement in advance of the
disruption is considered for ITER. The characteristic surface
temperature rise, which determines the material damage caused by the
power flux, can be estimated by the so-called "ablation-melting"
parameter  \~ Γ~Et.q.~/√t~t.q.~, where Γ~Et.q.~ is the energy flux
during the thermal quench and t~t.q.~ is its timescale. For the average
ITER disruption, the expected "ablation-melting" parameter will, thus,
be in the range 46--129 MJ m^-2^s^-1/2^, which is about a factor of 3-4
higher than that required to cause carbon ablation (35 MJ m^-2^s^-1/2^)
and tungsten melting (40 MJ m^-2^s^-1/2^) for ameliorated disruption and
64--322 MJ m^-2^s^-1/2^ for disruptions without amelioration actions.
While these values correspond by no means to a low energy flux, they are
typically about an order of magnitude lower than that expected under the
previous ITER assumptions (W~plasma~,~t.q.~ = 350 MJ, A~t.q.~ = 3
A~s.s.~, t~t.q.~ = 1 ms,  = 1054 MJ m^-2^s^-1/2^). As a consequence,
the expected divertor target lifetime under these "revised" most
frequent disruptions in ITER is significantly longer than previously
estimated.

While the new estimates alleviate the problem of energy deposition by
disruption at the ITER divertor and its lifetime, they lead to
significant energy flux reaching components (Be-clad blanket modules) on
which no significant disruptive energy flux was expected previously.

More details on the measurements and analysis can be found in "Expected
energy fluxes onto ITER Plasma Facing Components during disruption
thermal quenches from multi-machine data comparisons", A. Loarte, et
al., Paper IT/P3-34, Proc. 20th IAEA conference, Villamoura, Portugal,
2004.

4.2 -- 'Disruption mitigation studies in JT-60U', Kawano

It is well known that relativistic runaway electrons can be generated
during a tokamak disruption, and the plasma facing components would be
damaged if the localized and intense irradiation of runaway electrons
occurs. In order to mitigate the post-disruption runaway electrons with
good controllability, their characteristics with impurity pellet
injection were investigated for the first time using the JT-60U tokamak
device. A clear deposition of impurity neon ice pellets was observed in
the post-disruption runaway plasma. The pellet ablation was attributed
to the energy deposition of relativistic runaway electrons in the
pellet. A high normalized electron density was stably obtained with
n~e~^bar^/n^GW^ \~2.2. Prompt exhaust of runaway electrons and reduction
of runaway plasma current without large amplitude MHD activities were
observed. One possible explanation for the basic behavior of runaway
plasma current is that it follows the balance of avalanche generation of
runaway electrons and slowing down predicted by the Andersson-Helander
model, including the combined effect of collisional pitch angle
scattering and synchrotron radiation. Our results suggested that the
impurity pellet injection reduced the energy of runaway electrons in a
stepwise manner.

On the other hand, it has been pointed out that the current quench time
is extended by the appearance of runaway electrons. This fact suggests
that runaway electrons can be used to mitigate or avoid the current
quench when the safe and reliable control of runaway electrons is
available. Taking such a new point of view, an experiment for avoiding a
rapid current quench by allowing runaway electrons has been carried out.
An impurity neon pellet was injected as a killer pellet into an Ohmic
discharge during the period with the ECRF injection. Here, the current
quench was avoided presumably due to relatively high electron
temperature just after the pellet injection. While the electron
temperature decreased afterwards, runaway electrons were generated.
These runaway electrons reinforced the discharge to survive against the
low electron temperature of less than several tens eV and additional
impurity neon pellet injection. Thus the robust discharge was obtained,
and the plasma current was maintained and terminated as programmed. This
result indicates that the new concept for mitigation and avoidance of
current quench by runaway electrons is attractive.

4.3 -- 'DIII-D & Alcator C-Mod disruption mitigation', D. Whyte

Experiments show that the gas delivery rate is important for runaway
suppression and overall mitigation effectiveness. The gas must deliver
electrons (free + bound) in a time less than the current quench time.
The multi-device / multi-jet experiments will be needed to separate this
effect from any gas pressure scaling for edge penetration. Complementary
to the experimental work, realistic modelling of gas shock, including
friction, for gas delivery down the tubes is in progress.

Fast imaging shows that the gas/impurity interaction with the plasma is
complex. The toroidal and poloidal distribution of impurities and their
associated radiation will need to be better measured to extrapolate
radiation uniformity. Both experiments and models suggest that MHD plays
a role in final particle "penetration" to the center, without producing
large heat fluxes to the walls. Robust low-order MHD mixing lessens gas
jet requirements for ITER, but requires carefully benchmarked MHD
modelling, which has now started using the 3-D NIMROD code.

First experiments showed that gas jet delivery "worked" in C-Mod with
its high absolute plasma and magnetic pressure. Tailoring of the gas
species mixture can optimize thermal- and current-quench times.

4.4 -- 'Plasma shut-down with fast impurity puff on ASDEX Upgrade', S.
Günter

Neon gas puffing is routinely used to mitigate disruptions on ASDEX
Upgrade, triggered by a locked mode signal. Around 4.5x10^21^ atoms
(50-100% of the fuel inventory) are injected in 2ms. The neon does not
penetrate into the core plasma on the time‑scale of the thermal quench,
and builds up a cooling mantle at the edge (ρ\>0.8). Subsequent (1-2ms
later) narrowing of the core temperature profile and an order of
magnitude density increase triggers large MHD modes and disrupts the
plasma. Halo current forces are robustly mitigated with neon puffing,
decreasing with increasing neon puff pressure, but the impact on power
loads is still under investigation.

4.5 -- 'Disruption Mitigation at JET', P. deVries

Over the past years various disruption mitigation experiments have been
carried out. Gas injection has been applied to mitigate runaway
generation and pre-empt the current quench. However, it was found that
more gas and faster injection was required. In the 2004/5 shutdown a new
gas injection valve has been installed at JET, dedicated to disruption
mitigation. This fast gas valve, has a operating pressure between
5-35Bar and a response time of 0.5ms. Presently, the valve is connected
to the main chamber via a 4m long pipe, reducing its injection speed to
10Bar-litres in 100ms. Fast mitigation experiments require it to be
mounted closer to the vessel. For the coming campaigns several
experiments are planned to test the properties of the valve and the
effect of the gas injection onto the plasma and vessel conditions.

4.6 -- 'Far SOL ELM ion energies in Far SOL ELM ion energies in JET', R.
Pitts

Using a retarding field analyser (RFA) probe reciprocating into the JET
SOL at the top LFS of the poloidal cross-section, ion currents due to
Type I ELMs have been detected in a limited set of pure hydrogen
discharges at low I~p~, B~ϕ~. The low electrical bandwidth of the probe
and its associated electronics prevents conventional RFA voltage
scanning on the ELM timescale and so the energy of the detected ions is
inferred by fixing the internal ion retarding bias at the highest
permitted value and combining a simple model of RFA function with
predicted values for T~e~, T~i~ and n~e~ in the ELM filament. These
values are obtained using a new transient model for parallel energy loss
from the filament which requires as input the ELM plasma parameters at
the point of filament formation and the radial speed of filament
propagation. The latter has been characterised experimentally in the
past at JET, but the point of ELM filament origin is still a matter of
speculation. To account for this, experimentally observed "mid-pedestal"
values of T~i~, T~e~ and n~e~ are used and a range of expected values at
the RFA location computed for a ELM propagation distances corresponding
to filament creation at the separatrix or at the top of the pedestal.
The result is a predicted RFA ion current, representing the average
value carrried by the ELM, which is in very good agreement with the
observed values, bounded by the upper and lower limits corresponding to
the spatial uncertainty in the ELM filament origin. A sensitivity test
to radial propagation speed shows clearly that whilst a somewhat higher
value to that assumed on the basis of previous measurements (v~ELM~ =
0.6 kms^-1^) must also be considered compatible with the data, much
lower values cannot. The transient model predicts that T~e~ falls much
faster than T~i~ as the ELM filament propagates, an observation
compatible both with the RFA data and earlier direct measurements of
T~e~ on the ELM timescale in the JET far SOL. The fact that the model is
based on parallel loss rates determined by sheath boundary conditions
requires that it be seen as a description of the filament as propagating
in the SOL (as in the simple blob or plasmoid theory) but connected to
the divertor targets. The good agreement between model and experiment
lends considerable support to this picture. The model has also been
applied to estimate ELM ion impact energies on the ITER first wall at
the outboard midplane for the Q~DT~ = 10 Type I ELM reference scenario,
finding E~ion~ \~ 1.1 keV and thus considerably above even the tungsten
sputtering threshold. It also predicts \~8% of the ELM expelled energy
reaching the outboard limiters.

The presentation was based on two recently submitted JET papers, both
currently under review: R.A. Pitts, W. Fundamenski et al, "Far SOL ELM
ion energies in JET", submitted to Nucl. Fusion and W. Fundamenski and
R. A. Pitts, "A parallel transport model of tokamak power exhaust
transients", submitted to PPCF.

4.7 -- 'Heat and particle flux in the SOL and to the wall', A. Herrmann

Experimental data from MAST, DIII-D and AUG on filaments and ELM
extension in the SOL were presented. The observed behaviour of the
filaments can be described phenomenologically as field aligned
structures fed in the outer mid-plane with energy and/or particles. Most
of the ELM energy is deposited continuously in the strike point region
and only a fraction of the midplane loss is transported to the inner
wall. Fast Thomson scattering and Beam emission spectroscopy reveals the
origin of the filaments in the separatrix region. The radial velocity of
the filaments as measured by reflectometry, turbulence measurements
(Langmuir probes) and measurement of the delay between an ELM marker
(H~α~ ) and the onset of the ion saturation current at reciprocating
probes is below 1km/s, decelerated with radius at DIII-D and accelerated
at MAST and AUG. The filaments are rotating toroidally and/or
poloidally. Applying the velocity of toroidal rotation from
CXRS-measurements in the pedestal a toroidal mode number of about 10-15
can be deduced for AUG and MAST.

The extension of the filaments into the SOL is measured with
reciprocating probes, combined with thermography at AUG. The resulting
e-folding length for the heat flux and the ion saturation current is a
few centimetres. The inter ELM decay length tends to be shorter compared
to the ELM decay length which increases with decreasing density
(DIII-D). This results in an increasing contribution of ELMs to the
first wall ion flux (80% at low densities, DIII-D). The magnetic field
strength effects the variation (scatter) of the ELM size in the SOL
rather than the e-folding length (AUG, MAST). The interaction of
filaments with outboard limiters at AUG is very local with a few
centimetres poloidal width. Its contribution to the ELM power balance is
about 1%. The convective energy transport by filaments lasting about 100
μs is comparable to the mid-plane ELM energy loss.

4.8 -- 'ELM flux to the first walls on JT-60U', N. Asakura

ELM plasma transport in the SOL and divertor of JT-60U was analyzed
using an extensive array of reciprocating and target Langmuir probes.
Large peaks in particle flux propagate radially to the first wall at
1.3 -- 2.5 km/s (faster than in MAST, DIII‑D and JET), arriving in less
than the parallel transit time to the outer target, and deposit their
associated heat load locally (over a few cm).

4.9 -- 'Intermittent Transport of SOL/Divertor Plasmas in JT-60U', N.
Ohno

The role of fluctuations in causing particle transport in the far SOL
(∆r \~60-70mm) in JT-60U was explored. Fluctuations are found to be 5-10
times larger during ELMy H‑mode compared to L-mode and 4-5 times higher
at the outboard mid-plane compared to the X‑point.

4.10 - 'Summary of first wall loadings discussion', G. Counsell

Our knowledge of disruption power loads to the divertor has advanced
sufficiently that an update to the ITER physics design guidelines is
sensible. Data from most divertor tokamaks now show that the target heat
flux width broadens by a factor of 5-10 during the thermal quench,
larger than previous estimates, and that the plasma energy at the
thermal quench is a factor 2-4 lower than that during the full
performance phase of the discharge for most disruptions. There is
however still a need to quantify the impact of vapour shielding on
target heat loads during disruptions and the extent of heat loads on the
upper divertor. Our understanding of first wall loads in unmitigated
disruptions is much less developed, with data being exceptionally sparse
and ill conditioned. All that was generally known is what fraction of
disruption energy does not arrive at the divertor but much more work is
needed before any useful physics guidelines for the distribution of this
missing energy on other areas can be developed. Until this data becomes
available, estimates for the lifetime of first wall components in ITER
will be subject to large uncertainties.

On disruption mitigation, impurity injection (either as cryogenic
pellets or through gas puffing) had been demonstrated to provide a
reliable means of mitigating the impact of the current quench, both
runaways and halo currents, on several devices. The impurity injection
rate is important for mitigation, it being necessary to deliver
sufficient electrons (bound and free) to the plasma in a time less than
the current quench time. With gas puff injection, mixing light and heavy
species gases can help by increasing the effective flow velocity of the
heavy species and tailoring the mix could help optimise the thermal and
current quench times. The interaction of impurity puffing with the
plasma is complex with low order MHD appearing to play a role in
penetration of impurities to the core, which might impact on the
toroidal and poloidal uniformity of radiation on the first wall during
the mitigation. It was noted that impurity gas puffing, which dominated
the session, is only one possible technique and it is likely that a
range of systems might need to be employed (including for example killer
pellets) to handle different types of disruption. It was recommended
that a working group be established to develop a list of high priority
research items in this area over the next 2 years, which would include
further assessment of key issues such as the impact of MHD activity in
the plasma during the mitigation and the possibility that radiation flux
alone to the first wall could cause melting for some classes of
mitigated disruption.

On ELMs, it was agreed that there had been significant advances in our
understanding of first wall loads, including a deeper understanding of
their origin and structure provided by observations of filamentary
structures (and associated theories) on most machines. Desposition of
ELM energy onto radially protruding first wall components (such as
poloidal limiters) has now been observed on many machines (at least 25%
of W~ELM~ for large type I ELMs), although the poloidal and toroidal
localisation of this energy and it's temporal behaviour are not well
understood. Extrapolations for the limiter power load during type I ELMs
in ITER indicate potentially large melting and erosion rates for some
credible deposition scenarios. However, it was recognised that divertor
power loads may provide the 0^th^ order limit on allowable ELM size and
might actually necessitate operation in 'small' ELMs regimes (either
naturally small type I ELMs, type II/III, QH etc., or actively
mitigated, such as by rapid pellets or magnetic perturbation). It was
likely that limiter erosion would not represent a significant problem in
small ELM regimes, however even less was known about energy deposition
outside the divertor in these regimes and study was required, both for
the first wall and upper divertor.

As a result of the discussion on first wall loads, it was agreed that
the Divertor and SOL ITPA Group would aim to produce separate position
papers for ITER on the three main topics in this session (Disruption
power loads, ELM power loads and disruption mitigation strategies).

**Session 5: 'Next ITPA SOL/divertor meeting'**

**Location:** Several locations are being considered for the next 2
meetings. The primary choice for the meeting following this (so #7) is
to have it before the abstract selection meeting for PSI17. This will
likely be at the end of January or beginning of February in China. It is
possible it would be elsewhere but that will be up to the PSI chairman
-- Jiangang Li. Having the meeting in conjunction with the abstract
selection meeting makes it easier for the 5-6 people who are members of
both committees to attend.

Assuming that the 7^th^ meeting is in China, then the time of the 8^th^
meeting would be best in July. The choice from the point of balance
would be to have the meeting in Japan as we have not been there in a
while. However, that would mean 2 meetings in Asia in a row, not to
mention that there will be a PSI meeting in May in China -- so 3 trips
to Asia! Alternatively we could try and have the 8^th^ meeting back in
the US or Europe. A third possibility is to wait till the fall and have
the 8^th^ meeting to coincide with IAEA but this will be back in China!

**Potential Topics for the next meeting** -- A number were discussed and
found support:

1\) D/T retention in surfaces (Long-pulse, saturation....., processes
involved)

2\) Wall conditioning (characterizing the effect on confinement, methods
of steady-state replenishment compatible with ITER conditions)

3\) Startup studies (flux consumption, operation on outside limiter)

4\) Flow characterization

5\) melt layer stability and vapor shielding

6\) spectroscopy of hydrocarbons in low-T plasmas

**Session 6: 'Mixed material effects'**

6.1 - 'Results on alloy formation of Be and W'**,** J. Roth

The interaction of Be with W to form low melting point W-beryllides was
performed in a collaboration of IPP Garching and PISCES, UCSD in three
procedures: The interaction of a 235 nm W layers deposited onto
polycrystalline Be, the slow evaporation of 5 nm Be onto W, and the
exposure of W to a Be-seeded D-plasma.

1.  The results show that W-beryllide phases are real, and bear the
    potential of a major malfunction: At 1070 K W-beryllide phases can
    form which melt below 1700 K. For W on Be this phase has clearly
    been identified as Be~12~W through ion beam and XPS analysis. The
    alloy formation appears to be limited by the diffusion of Be in the
    alloy.

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1.  However, under typical ITER conditions (0.1% Be in incident D flux)
    only a thin Be~2~W layer will form, excess Be will
    sputter/reflect/evaporate and deposit elsewhere. Studies using
    higher Be plasma concentrations (up to 1% ) are underway in
    PISCES-B. The stability of the beryllide layer during ELM-like
    transient heat loads will be investigated in PISCES-B

6.2 -- 'PISCES Be/C mixed materials studies'**,** R. Doerner

Mixed material investigations of the Be-C system in the PISCES-B
facility were described. Be-rich surface layers form on plasma-exposed C
samples and lead to co-deposits that are composed almost entirely of Be
and D. Although the level of co-deposition of Be and D is on the same
order as that expected from co-deposition of C and D, the D retained in
Be is released thermally at much lower temperature. When C is included
in the co-deposits an amount of the D retained in the co-deposits is
trapped much more strongly and cannot be easily removed by baking.

6.4 - 'The IPP dual beam experiment -- first results on simultaneous
irradiation of tungsten by deuterium and carbon'**,** K. Krieger

Plasma facing surfaces in tokamaks are generally subject to impact of a
mixture of fuel ions and impurities. To study the basic processes
involved in multi-species bombardment, a dual beam experiment has been
commissioned at IPP Garching. First results are presented on erosion
rates and fuel retention under simultaneous irradiation of tungsten at
room temperature with deuterium and carbon. In contract to results
obtained for a 25% fraction of C, irradiation of C with D and a 5%
fraction of C yields continuous erosion of W enhanced significantly over
pure D bombardment. Carbon is implanted in the surface and evolving to a
mixed surface layer with stationary thickness and composition. Although
the principal processes have been identified, the new results are a
challenge for the existing models.

6.4 'Mixed materials discussion and conclusions'. J. Roth:

The following discussion of Mixed Material Session was structured
according to the three main expected material interactions and their
potential specific problems:

**Be-W system**: This system bears the potential for major malfunction.
Layers may form with low melting point (\< 1700°C) and may have bad
thermal contact to substrate. The process must be discussed under the
most likely ITER scenario, i.e. the exposure of W to a Be-seeded plasma,
and all processes such as deposition, reflection, re-erosion and
evaporation have to be taken into account to assess the failure
potential. Disruptions may deposit up to 10 kg of evaporated Be on W
divertor structures. Be-W system needs extensive investigation, appears
most critical.

**Be-C system**: Experiments in PISCES show, that deposited Be reduces
the chemical erosion of C. Still, for interpretation of results in JET,
the simultaneous Be and C deposition from the D-plasma needs to be
investigated.

Further needed investigations are: What is influence on T retention? The
Be/C mixture exhibits similar D co-deposition at 300 K, but D release
occurs at much lower temperatures than for carbon.

**W-C system**: W erosion in present devices (ASDEX Upgrade) is enhanced
by C impurity ions in the plasma

As W is eroded by C ions, the C is removed by D ions leaving the clean W
surface exposed. W-carbide formation occurs at temperatures above 700°C
and is well understood. Deviations from thermo-dynamical phase diagrams
occur due to the highly non-equilibrium conditions.

In the discussion it was pointed out that a much larger variety of
mixing compositions may occur during the history of different fusion
experiments which lead to hardly reversible wall conditions. Much larger
varieties of mixing conditions should be studied in ion beam
experiments, plasma simulators and fusion experiments, such as JET,
ASDEX Upgrade etc.

Under clean conditions, still seed impurities for radiation are present
and may contribute to material modifications.

It was concluded that it seems impossible to cover all possible
combinations and simulate identical ITER divertor conditions. This
requires that test experiments need to be extensively simulated in
computer codes and the conditions be extrapolated using validated data.
The main pathway needs to be finally explored in the fusion experiments,
while laboratory experiments remain necessary to avoid major pitfalls.

**Session 7 : Estimates of the effects of Be operation in ITER. Leader -
A. Loarte**

This session reviewed the implications of the use of a Be wall in ITER,
including the lifetime of the wall, Be migration, implications of the
Be-wall for T-retention and removal in ITER and general considerations
on the operability of ITER associated with the main chamber being made
of Be.

7.1. Introduction. A. Loarte

This talk reviewed the physics basis which lead to the decision of a Be
wall in ITER and discussed the needs for a re-evaluation of this
decision in view of the new findings in tokamaks since 1998. Be was
chosen for first wall material in ITER because its low Z compatibility
with a wide operating range and low T-retention under the following
assumptions:

a\) The flux of ions on the wall is determined by diffusive anomalous
transport and C--X as modelled by B2-Eirene. The typical ratio of
divertor ion flux to main wall ion flux is larger than 100. This is
associated with a low Be erosion of the wall during normal operation of
the device and small associated T-retention.

b\) There are no significant power loads (compared to divertor) on the
wall during ELMs, as the distance from the separatrix to the first
Be-PFC elements is more than 7 e-folding lengths for the power energy
flux during ELMs (no significant broadening of the ELM power flux was
assumed).

c\) There are only small power loads (compared to divertor) on the wall
during disruptions, as the distance from the separatrix to the first
Be-PFC elements is more than 2 e-folding lengths for the power energy
flux during disruptions (only a factor of 3 broadening of the disruption
power flux was assumed).

d\) Issues related to material-mixing, its influence on T-retention and
cleaning techniques were not considered in detail.

7.2. Be wall erosion in steady-state (Kallenbach, Philipps, Federici,
Roth).

Estimates of the average Be wall erosion presented were based on
physical sputtering of Be by ion impact and C-X neutrals from modelling
results with B2-Eirene and various experimental extrapolations with the
following results :

-   B2-Eirene extrapolations. Depending on modelling assumption on the
    plasma in the far SOL the expected Be erosion is 4.5 10^21^ s^-1^
    (Federici) to 8.0 10^21^ s^-1^ (Roth).

-   Experimental extrapolations range from 2.0 10^21^ s^-1^ (Philipps)
    to 1.2 10^23^ s^-1^ (Kallenbach) depending on whether the presently
    measured wall fluxes are assumed to be similar to those in JET in
    absolute amount or whether an empirical relation which fits the JET
    and ASDEX Upgrade density and wall fluxes is used.

Typical rates taking into account that the main chamber Be will probably
form BeO are a factor of 2 lower. The resolution of this issue relies on
the experimental scaling of the ion fluxes on the wall and whether there
is a favourable or unfavourable size scaling of the anomalous particle
flux density or not. If the total flux on the wall in ITER is similar to
today's experiment, the expected erosion rate of \~ 40 nm/discharge
allows ITER operation with a Be wall for its expected operating life
(Federici). On the contrary if the anomalous ion flux density in ITER is
similar to today's experiment, Be erosion will be in the range of \~1
μm/discharge and the Be wall lifetime limited to several thousand full
performance discharges (Kallenbach). The above values represent average
values of Be erosion of main wall components, the local erosion of
components which are closer to the plasma such as the upper x-point
modules and limiters could be significantly higher than these average
values.

7.3. Be wall erosion under transients (Herrmann, Whyte).

Estimates of the Be erosion under ELMs based on extrapolations from
measurements in ASDEX Upgrade indicate that for ELM loads that do not
cause significant target erosion (sublimation of C or melting of W),
they will not cause Be melting of main wall elements, even of the upper
X-point modules (Herrmann). These estimates are in agreement with
estimates done by A. Loarte and presented at the PSI 2004 in Portland.
However, significant melting of the Be limiter could occur in these
conditions. The situation is much less clear for disruption loads in
which significant melting of the Be wall and of the Be limiter may occur
even if the energy is spread over a large area of the wall during the
thermal quench (Herrmann).

This will be indeed the case for mitigated disruption by impurity
injection and VDEs (Whyte). Ideal rapid, quasi-uniform radiation flash
desired for disruption mitigation can lead to melting of the \~500 m^2^
beryllium wall. Q=10 plasma termination produces a large-area thin melt
layer (\~10's microns) that remains molten during thermal quench \~
0.2 - 2 ms. Because the layers melt simultaneously with the dissipation
of the plasma thermal energy, the layer is exposed to a radial inward,
mobilizing JxB force resulting from induced toroidal eddy currents
caused by the expulsion of the diamagnetic toroidal flux through the
highly electrically conducting Be wall. The body acceleration is \~ 100
g and the layer displacement is found to be \> mm with terminal velocity
\~ m/s. Surface tension acts as a stabilizing force, but for nominal
tile size is insufficient to balance JxB (\~ 1mm surface "dimples" may
work). Since little plasma is present to stop the melt layer, it is
expected that 10-50 kg of Be will likely "splash" onto other main-wall
surfaces and down into the divertor by gravity. Similar quantities of Be
loss were found for an unmitigated VDE with thicker melt layers (0.5 mm)
spread over a smaller area (\~30 m^2^). The mobilized Be will likely be
in the form of droplet due to the surface tension and these will present
non-uniform surfaces, both at the location of melting and where it
splatters, for heat loading in subsequent discharges.

Finally, it was highlighted that concentrated energy deposition from
fast particles could be a problem for the Be wall in ITER, but no
estimates of these loads in ITER were considered (Herrmann). In view of
the results, developing regimes which are acceptable for the divertor
lifetime will probably lead to regimes which are acceptable for Be
erosion of the main wall with respect to ELMs. The open issues are the
ELM loads on the ITER Be limiter and the dynamics of the molten layer
during disruption, mitigated disruptions and VDEs. Experiments in
simulation facilities and tokamaks are necessary to address the issues
besides more refined estimates of the expected erosion.

7.4. Main wall eroded Be migration to the divertor and C/W migration
onto the Be wall in ITER (Philipps, Roth, Kukushkin, Coster).

Two approaches were presented to do these estimates: a) one based on the
experimental evidence, which indicates that most eroded material from
the main wall ends up at the inner divertor and that the main wall is a
net erosion zone; and b) modelling by B2-Eierene of the erosion
deposition balance. In the first estimate the main conclusion is that
most of the eroded Be will be transported to the divertor (mostly to the
inner one) and will deposit there in areas with direct plasma contact
(Philipps, Roth). Following this empirical picture, it is expected that
C and/or W escaping the divertor will only be found at the wall
transitorily and will finally migrate to the divertor (predominantly the
inner) were it will deposit permanently. Some deposition of C and W
could be expected in shadowed areas of the main wall but this is not
expected to be a major contributor to overall deposition. Modelling with
B2-Eirene provides a much more complicated picture for main
wall/divertor erosion and material migration following the studies by
Kukushkin and Coster. The divertors are found to be deposition zones for
the eroded material both from the divertor and the wall, except close to
the strike zones where significant carbon erosion can take place. The
main wall has areas which can be deposition dominated and erosion
dominated. Both modelling studies show material deposition from the
divertor (Carbon) on the LFS of the main wall and erosion (of Be) in the
HFS. The rate of carbon deposition is obviously affected by the rate of
re-erosion of the deposits. If this is large enough, the balance from
deposition to erosion and its location can change significantly in the
calculations. To clarify which features of this modelling are ITER
specific and which ones are associated with inaccuracies in the
modelling of SOL flows, which are seen to be present in present
experiments, and re-erosion of deposits requires a detailed comparison
of similar modelling results for actual experiments. Using markers to
determine with high precision the migration of materials in the
experiments is highly recommended as it facilitates the interpretation
of experiments and validation of models.

7.5. Implications of a Be wall on T retention and removal in ITER
(Philipps, Roth, Federici, Counsell).

The typical concentration of T in Be is very low, typically in the 1-5 %
level, which amounts to a retention in the range of 0.1-0.6
g-T/discharge depending on modelling assumptions (Philipps, Roth). This
amount is similar to that expected for carbon chemical erosion if the
target is assumed to be covered by Be following ERO modelling and simple
estimates (Roth, Counsell). However significant T release is expected if
these Be-co deposits are brought up to temperatures higher than 600 K
(Federici, Roth). This is likely to be the case at the ITER divertor, as
Be deposits only occurs in plasma wetted areas of PFCs (Federici, Roth).
If Be is expected to be in the form of BeO and co-deposit at the
divertor the retention could be larger. Estimates of the T retention on
BeO based on the incoming of oxygen due to the reference leak rate of
the ITER vessel or by purity of the initial material show that the
amount of T retained is similar or smaller than that associated with
pure Be for ITER operation excluding transients (Counsell). It is
expected that the T retained in the Be deposits will be easily removed
by energy deposition of the plasma itself which keeps the divertor
surface temperature high. If this is not the case, the present foreseen
techniques for T removal from C, such as photonic cleaning or oxidation
will be difficult to apply, if not impossible. In summary, for Be
erosion rates which are acceptable with the lifetime of the Be wall the
amount of T retained in ITER will be in the 0.1-0.6 g-T/discharge at
most. The deposits will be localised in areas where plasma exposure
should be effective to remove the deposited T. Oxygen and C content in
the layers makes the retention larger but still acceptable by control of
the divertor target temperature by plasma exposure within standard
ranges of ITER operation. Despite this positive picture, it is important
to highlight that none of the foreseen schemes to remove T from
hydrocarbon deposits seem applicable for Be and some T-removal technique
from Be deposits should be developed in case it is necessary. Detailed
analysis of Be-containing samples exposed to plasmas in tokamaks or
divertor simulators and dedicated experiments to remove the retained
fuel from such samples are required to progress further in this area.

7.6. Operability of ITER with a Be main wall/limiter including effect of
PFCs damage due to transients (Kallenbach, Counsell, Loarte, Federici,
Whyte).

The major issues in this area refer to the damage of the Be wall and the
effect of a damaged wall on ITER operations. Main Be wall erosion
between ELMs, although potentially a lifetime issue for the wall does
not seem to be a potential threat to the purity of ITER plasmas
(Counsell, Kallenbach). Damage by ELMs is likely to cause irregularities
on Be-wall components in the range of 10-100 μm which do not appear to
pose a danger for the integrity of the components (Counsell, Loarte,
Federici). These irregularities could, however, compromise the power
handling capabilities of the Be-PFCs, in particular of the limiter,
which is required to have a good power handling and will be exposed to
direct plasma contact during the ramp up and ramp down phases of the
discharge (Counsell, Federici). The expected effect of such occasional
Be melting of the first wall on ITER plasmas is the occurrence of
transient increases of the Be impurity concentration and decrease of the
fusion performance. Causing a radiative collapse of an ITER plasma by Be
influx requires about a 200 g influx of Be (Loarte). Depending on the
frequency of such influxes and the amount of Be that gets into the
plasma, the real consequences for the fusion performance of ITER are
obviously different. Disruptions (even if mitigated) and VDEs, on the
contrary, have the potential to cause serious damage to the Be wall
(Counsell, Whyte, Federici) and can cause several kgs of Be to be eroded
from the wall after every one of these events. In order to progress
further in this field and provide better estimates for ITER it is,
therefore, necessary to: a) collect experimental evidence of plasma
response to well know amounts of low Z solid impurity influxes (to
determine plasma resilience to the ELM Be-influxes in ITER, b) carry out
modelling and experiments of the Be-melt layer dynamics under ELM and
disruption loads and c) develop scenarios to ameliorate the effects of
disruptions and VDEs on the Be wall.

7.7. Status of the JET ITER-like wall project. V. Philipps.

Be blanks for the main wall, Be cladding, and new CFC tiles to be coated
with W for the ITER-like wall project have been tendered. The total
amount of Be tiles in the main chamber will be somewhat reduced due to
budget problems with ordering of the Be blanks. The possible use of Be
plasma-sprayed Inconel tiles between the inner wall guard limiters will
be decided soon. With respect to W coating of the CFC divertor tiles,
prototype coatings with 5 and 10μm and "thick" (\>200μm) W-layers on JET
Dunlop CFC are underway, to be tested in the fall under high heat load
conditions. A bulk W-tile is under construction for the horizontal
divertor tile (number 5, LBSRP tile). The project also includes upgrades
of diagnostics (spectroscopy, IR and deposition probes).

**Session 8: 'The ITER dome'**

8.1 - 'Original strategy for the dome and its physical functions', M.
Shimada

The present strategy of the dome is to investigate the physical
functions of the dome and pros and cons of its removal and to
investigate the consequences of possible dome removal on diagnostics and
other in-vessel components. Originally the dome was intended to reduce
the neutral backflow from the divertor to the main plasma to facilitate
the partial detachment for the enhancement of radiative cooling and
particle exhaust and to protect the liner and the pumping slot from the
plasma. However, much deposition of tritium, carbon and dust is expected
at the backside of the dome and the liner, where measurement and removal
are difficult. Furthermore, experimental and modeling studies suggest
that the shielding effect including the neutral-neutral collisions could
reduce the neutral backflow. Therefore a proposal of removing the dome
has been made to possibly facilitate the measurement and removal of
tritium, carbon and dust. For the improved argument of the benefit of
the dome removal, it is necessary to estimate the deposition rate on the
backside of the dome with consideration of the effect of parasitic
plasma and to estimate the deposition rate of dust below the pumping
slot.

8.3 - 'ITER calculations of divertor performance with & without the
dome', A. Kukushkin

In order to study the effect that the dome removal would have on the
plasma performance in ITER, a number of modeling runs was done with the
B2-Eirene code including the recent upgrade of the neutral particle
transport model (neutral-neutral collisions, improved molecular
reactions, Lyman radiation transport). The results indicate that the
most significant effect would be an increase of the required pumping
speed at the pump duct entrance by a factor 3, needed to compensate for
the enhanced neutral particle absorption by the plasma. Other aspects of
the dome removal, such as placing the divertor diagnostics, reduction of
neutron shielding, design of the pumping slots, and so on, must still be
thoroughly considered before the removal of the dome could be
recommended for the ITER design.

8.4 - 'Interpretive analysis of the C-Mod divertor for partially
detached plasmas: possible implications for the ITER dome', Lisgo

Tangential camera images in D~γ~ light have shown significant emission
from the C-Mod private flux region (PFR) for plasmas that are weakly
attached in the outer SOL (the inner SOL is almost always partially
detached), with the emitting region extending from the separatrix to
near the divertor floor. The transport processes generating this
extended plasma structure in the PFR are poorly understood, and may
result from the collisional nature of the C-Mod divertor. OSM-EIRENE
modeling suggests that the divertor neutral pressure is primarily
controlled by the PFR plasma in this instance, a result to consider when
discussing the physics implications of including a dome in the ITER PFR
(since the collisionality of the ITER divertor is also expected to be
high). The C-Mod study indicates that the neutral processes included in
the EIRENE neutral code are sufficient for first order estimates of
divertor neutral pressures in ITER, provided that a reasonably accurate
description of the ITER divertor plasma is available.

8.5 - 'Summaries of talks on the ITER dome', D. Whyte

Consideration is being given to removing the dome from the ITER divertor
design. This is motivated by the desire to inhibit possible tritium
codeposition and dust accumulation underneath the dome, and to ease
geometric access to the divertor for tritium removal by optical cleaning
techniques.

Originally the dome was primarily intended to reduce the neutral
backflow from the divertor to the main plasma, to facilitate partial
detachment, and to enhance particle exhaust. However both experimental
data (e.g. septum removal in JET) and the recent improvement of
neutral-neutral collisions in the B2-EIRENE modeling appears to have
removed these motivations. No significant change was found in divertor
parameters (peak heat load, radiative cooling) or upstream plasma
(density limit, impurity content) as a function of the absence or
presence of dome. The most significant effect would be an increase of
the pumping speed at the pump duct entrance by a factor 3, needed to
compensate for the enhanced neutral particle absorption by the plasma.
This pumping speed appears to be within the present capability in ITER.
The issue of increasing neutral leakage out of the divertor by removing
the dome was raised. It was pointed out that modeled neutral leakage
from the private-flux region (PFR) was very small, due to the much
larger ratio of size to neutral mean free path in ITER as compared to
present devices. Interpretative modeling of the C-Mod divertor, which
most closely approaches ITER divertor collisionality, showed that the
inclusion of neutral-neutral collisions was required to reproduce the
measured baffle pressure, generally validating the new B2-EIRENE results
for ITER. An additional insight from the C-Mod modeling was that the
presence of dense cold plasma in the PFR plasma largely determined the
baffle neutral pressure. Since the reasons for this PFR plasma are
poorly understood this raises questions about the implication of a dome
in the PFR of ITER.

Another important purpose of the dome, which evolved from its long-term
presence in the ITER design, is the protection of components associated
with pumping (liner, pump slot) and optical components of divertor
diagnostics. The most worrying implications of dome removal appear to be
divertor diagnostic losses: IR thermography of the target plates,
spatially resolved visible impurity spectroscopy, and the laser system
for measuring divertor plate erosion are all lost. The absence of these
diagnostics, particularly the IR required to measured target plate
temperature, could have a strong impact on divertor operation. A full
evaluation of the impact on diagnostics, including determining the
consequences to the machine operation and physics studies arising from
the reduced measurement capability, should be completed before a
decision to remove the dome is taken.

Furthermore, no detailed engineering design is yet available as to what
would replace the dome structure in ITER. For instance, the detailed
technical aspects of protecting the pumping slots, liner and lower floor
from plasma and neutrons were not yet available. Such a design is
required for a realistic assessment of the diagnostic implications, as
well as the implications for tritium removal,.

A further concern raised during the session was the level of certainty
about the presence of tritium codeposits and dust under the dome in
ITER. It was pointed out that present tokamak experience can be
misleading. Due to the fact that neutrals are in the fluid regime in the
ITER divertor, line-of-sight carbon deposition, which seems to dominate
deposition in present devices, will not occur in ITER. Dust formation
and transport is even less certain. It was also unclear if the simple
removal of the dome would ameliorate tritium retention or recovery,
since the exposed pump liner may serve the same purpose of trapping and
hiding tritium and dust. Again a more detailed alternative divertor
design seems required.

Conclusions

No consensus was reached on the balance of benefits vs. losses with
regard to the dome removal. While tritium and dust recovery are clearly
important issues to the operational viability of ITER, it remains
unclear if there is yet enough evidence to remove the dome based on
present data or modeling. The dome removal clearly has important
consequences to divertor diagnosis, so this must be balanced against the
tritium inventory considerations.

Suggestions for further work on this topic include:

-   Further studies on the origin and effect of the PFR plasma.

-   Initiating an engineering design as to what would replace the dome
    structure in ITER considering protection of the vessel and pumping
    slots.

-   Assessment of the neutron shielding issues with the dome removal.

-   A full evaluation of the impact on diagnostics, including
    determining the consequences to the machine operation and physics
    studies arising from the reduced measurement capability.

-   The effect of ITER neutral collisionality on carbon atom and
    molecule deposition patterns with and without the dome (likely
    coupled to PRF plasma studies).

-   Empirical assessments on carbon/deuterium/dust accumulation behind
    "dome-like" structures in present devices.

# Appendix A - Agenda

**1^st^ Day: July 4^th^ (Mon.)**

![](media/image1.pct){width="6.295833333333333in"
height="3.8131944444444446in"}

**2nd Day: July 5^th^ (Thu.)**
![](media/image2.pct){width="5.7444444444444445in"
height="3.9159722222222224in"}

**3rd Day: July 6^th^ (Wed.)**

![](media/image3.pct){width="6.461805555555555in" height="4.41875in"}

**4th Day: July 7^th^ (Thu.)**

![](media/image4.pct){width="5.103472222222222in"
height="1.2493055555555554in"}

## **Appendix B - Meeting attendees** 

## Total 45 : EU(18)+JP (6)+US(11)+CN(2)+RF(2)+ITER(6)

> **Attendee Institute e-mail**
>
> ***EU(18):***
>
> C. Hidalgo (CIEMAT) carlos.hidalgo@ciemat.es
>
> F. Tabares (CIEMAT) tabares@ciemat.es
>
> V. Philipps (FZJ) v.philipps@fz-juelich.de
>
> A. Kreter (FZJ) a.kreter@fz-juelich.de
>
> O. Ogorodnikova (FZJ) o.ogorodnikova@fz-juelich.de
>
> K. Krieger (IPP-Garching) krieger@ipp.mpg.de
>
> V. Rohde (IPP-Garching) rohde@ipp.mpg.de
>
> A. Kallenbach (IPP-Garching) Arne.Kallenbach@ipp.mpg.de
>
> J. Roth (IPP-Garching) roth@ipp.mpg.de
>
> A. Herrmann (IPP-Garching) albrecht.herrmann@ipp.mpg.de
>
> D. Coster (IPP-Garching) dpc@ipp.mpg.de
>
> C. Grisolia (CEA) christian.grisolia@cea.fr
>
> E. Tsitrone (CEA) emmanuelle.tsitrone@cea.fr
>
> X. Bonnin (CEA) bonnin@limhp.univ-paris13.fr
>
> G. Counsell (UKAEA) glenn.counsell@ukaea.org.uk
>
> R. Pitts (CRPP) Richard.Pitts@epfl.ch
>
> A. Kleyn (FOM) kleyn@rijnh.nl
>
> A. Loarte (EFDA-Garching) loartea@ipp.mpg.de
>
> ***JP (6):***
>
> Nobuyuki Asakura (JAERI) asakuran@fusion.naka.jaeri.go.jp
>
> Naoko Ashikawa (NIFS) ashikawa@lhd.nifs.ac.jp
>
> Tomohide Nakano (JAERI) nakanot@fusion.naka.jaeri.go.jp
>
> Noriyasu Ohno (Nagoya Univ.) ohno@ees.nagoya-u.ac.jp
>
> Seio Sengoku (JAERI) <sengokus@fusion.naka.jaeri.go.jp>
>
> Yoshio Ueda (Osaka Univ.) <yueda@eie.eng.osaka-u.ac.jp>
>
> ***US(11):***
>
> Bruce Lipschultz (MIT) blip@psfc.mit.edu
>
> Peter Stangeby (Univ. Toronto/GA) stangeby@fusion.gat.com
>
> Sergei Krashnenikov (UCSD) skrash@mechanics.ucsd.edu
>
> Dennis Whyte, (Univ. Wisconsin) whyte@engr.wisc.edu
>
> Max Festenmacher (GA) Max.Fenstermacher@gat.com
>
> George Tynan (UCSD) gtynan@ucsd.edu
>
> Russ Doerner (UCSD) rdoerner@ucsd.edu
>
> Steve Lisgo (Univ. Toronto) steven@starfire.utias.utoronto.ca
>
> Rajesh Maingi (PPPL) rmaingi@pppl.gov
>
> Mathias Groth (GA) <groth@fusion.gat.com>
>
> Vlad Soukhanovskii (LLNL) <vlad@pppl.gov>
>
> ***CN (2):***
>
> Yu Yang (IPP) <yangyu@ipp.ac.cn>
>
> Jiansheng Hu (IPP) <hujs@ipp.ac.cn>
>
> ***RF(2)***
>
> Valery Kurnaev (MEPhI, State Univ.) kurnaev@plasma.mephi.ru
>
> Gennady Kirnev (Kurchatov Inst.) kirnev@nfi.kiae.ru
>
> ***ITER IT (6):***
>
> Gianfranco Federici (ITER IT) federig@itereu.de
>
> Kukushkin Andrei (ITER IT) <kukusha@itereu.de>
>
> Michiya Shimada (ITER IT) shimadm@itergps.naka.jaeri.go.jp
>
> Kiyoshi Itami (ITER IT) <itamik@fusion.naka.jaeri.go.jp>
>
> Alan Costley (ITER IT) costlea@itergps.naka.jaeri.go.jp
>
> Horst D. Pacher (ITER IT/INRS-EMT) pacher@emt.inrs.ca