Fuel Inventory in Shadowed Areas of the JET Divertors
M. Rubela, J.P. Coadb, D. Holec and JET EFDA ContributorsaAlfvén Laboratory, Royal Institute of Technology (KTH), Association EURATOM-VR, 100 44
Stockholm, Swedenb
Culham Science Centre, EURATOM / UKAEA Fusion Association, Abingdon, Oxon OX14
3DB, United KingdomcSchool of Mathematical and Physical Sciences, Accelerator Laboratory, University of Sussex,
BN1 9QH Brighton, United Kingdom

Components of JET divertors have been examined to in order to assess the fuel retention inareas shadowed from the direct plasma impact: castellation groves in beryllium blocks, gaps separatingtiles of Mk-I, and the interior of the septum module in the Mk-II Gas Box (GB) structure. The results
show that: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) in thesegaps the fuel inventory exceeds that on plasma facing surfaces by up to a factor of 2; (iii) in the groovesof castellation the fuel content is less than 3.0 % of that found on top surfaces; (iv) fuel retention in the
septum of Mk-IIGB is insignificant in comparison to that detected in the shadowed region in the innerdivertor corner. Implications of these results for a next-step device are addressed.
Keywords: fuel inventory, material migration, carbon, beryllium, JET
I. Introduction
Safety and economy of a fusion reactor
operation are the driving forces for theassessment of fuel accumulation in plasma facing
components (PFCs). The issue has been known
for long time, but its full seriousness has beenrealised following the D-T campaigns carried out
in tokamaks with a carbon-based first wall: JET[1] and TFTR [2]. It has also been recognised thatthe divertor geometry and the related power
deposition profiles have a strong impact on thetransport of eroded material and its co-deposition
together with fuel species. For that reason, an
important mission of the JET tokamak is to
optimise the wall and divertor geometry and totest a variety of plasma operation scenarios.
---------------------------*

See Appendix to the paper by J. Pamela et al., Fusion

Energy 2004 (Proc. 20th Int. Conf., Vilamoura 2004)
IAEA, Vienna 2004.

Over the last decade the main chamber wall anddivertor of JET have been restructured several
times. Since 1994 it has been consecutively
operated with the following divertors: Mk-I withcarbon fibre composite (CFC) and then withberyllium tiles and, three Mk-II structures with
CFC tiles: Mk-IIA, Mk-II Gas-Box (GB) and
Mk-II Septum Replacement Plate (SRP). Tiles in
the main chamber wall are made with carbon andthe wall is regularly coated by an evaporated
beryllium layer. When the fuel retention andmaterial transport studies are considered, thesimultaneous use of carbon and beryllium is very
important for at least two reasons: (i) these twomaterials are planned to be used also in ITER; (ii)it helps to distinguish chemical and physicalprocesses in wall erosion and to conclude on theirimpact on fuel inventory.
Following the full D-T campaign at JET
with the Mk-IIA structure the most pronounced

accumulation was found in the inner divertor
corner, in particular, in remote areas shadowed
from the direct plasma line of sight [3-5]. After
this experience detailed studies have been
undertaken to assess the material transport andfuel retention to shadowed regions of other JET
divertors. The emphasis was given to studies ofre-deposition and fuel retention in certain
construction features similar to those foreseen for
ITER: castellated tiles and dome. The issue ofmassive co-deposition in such areas is of crucial
importance for ITER because they are difficult toaccess by cleaning methods for fuel removal.
The castellated structure of tiles in the
ITER divertor is thought to be the best solution toensure the thermo-mechanical durability andintegrity of materials under high heat flux loads,
especially when considering the use of metals(tungsten and beryllium). Until recently, the Mk-I
divertor has been the only large-scale castellated
structure used in fusion experiments. Therefore,
analysis of deposition in castellated grooves andin gaps between tiles of Mk-I may contribute to
the assessment and improved understanding ofmaterial transport and mixing in an environment
containing both C and Be.
Strong emphasis given to the analysis
of surfaces inside the gas box module of the
Mk-IIGB divertor is justified by the fact that its
structure has been, so far, the closest to theplanned geometry of the dome in the ITER
divertor. The inner part of septum has had alarge surface area shadowed from the direct
impact of confined plasma. Therefore, this region
could be deemed as a potential trap forco-deposition associated with significant fuelinventory.
The main objective of this paper is toprovide an overview of fuel inventory in
structures resembling those to be implemented in
ITER: (i) castellation which was used in Mk-I
and (ii) the interior of the septum. The deposition
pattern in the inner corner of Mk-IIGB and a brief

comparison of deposition in two Mk-II divertors(Mk-IIA and Mk-GB) will also be given.
II. Experimental
II.A. Mk-I Divertor
Mk-I was a water-cooled structurecomposed of small roof-shaped tiles: 72 mm long
poloidally, 30-40 mm wide toroidally, 50 mm
high and, as shown in Fig. 1, separated by gaps up
to 10 mm wide. As a result of roof-shaping andarrangement of the tiles with respect to themagnetic field lines a lower part of a given tile
was shadowed by the upper edge of the adjacentone. The divertor was operated first for a total of
59661 s with CFC and then for 20303 s withcastellated (6x6 mm with 6 mm deep and 0.6 mm
wide groove) beryllium blocks. Images in Fig. 2
feature the structure with CFC (a) and Be (b)tiles.

Fig. 1.Arrangement of tiles in the Mk-I divertor.

Fig. 2. Inner part of the Mk-I divertor with
CFC (a) and castellated Be (b) tiles

II. B. Mk-IIGB Divertorµm to approximately 10 µm, respectively.

Fig. 3 a shows the cross-section of
Mk-IIGB and the most typical position of strike
points on the vertical divertor plates. In Fig. 3 b
the scheme of the septum module is depicted.
Tile 5 and divider plates were positioned
toroidally, whilst the three support plates wereoriented poloidally. Experimental campaigns
with that divertor lasted for about 18 months in
the period 1999-2001. The total operation timewas 193291 s including 94299 s of X-point
plasma.

III. Results and Discussion
III. A. Deposition in Gaps and in Castellation
III. A. 1. Mk-I Divertor with CFC Tiles
Fig. 4 shows the appearance of the CFC
tiles after the long-term operation. There is asignificant deposition on the plasma facing
surfaces and also on the sides, i.e. on surfaces
located in gaps between the tiles.

Plasma facing surfaces
Side surfaces

Plasma facing surfaces
Fig. 3. Cross-section of the Mk-II Gas Box
divertor (a) and the structure of septum (b).
II. C. Analysis methods
Tiles retrieved from the torus were
examined ex-situ by means of ion beam analysis(IBA) techniques. Nuclear reaction analysis(NRA) was used to quantify and map surface
contents of deuterium [3He(d,p)α], beryllium[3He(9Be,p)11B] and carbon 12 and 13 isotopes:[3He(12C,p)14N] and [3He(13C,p)15N]. Most of theD, Be and C analyses were carried out with a 2.5
MeV 3He+ beam, but to assess the thickness of
deuterium-containing deposits, the 3He beamenergy was occasionally scanned in the range
0.7–3 MeV giving the information depth from 1.3

Fig. 4. CFC tiles of the Mk-I divertor: generalview (a), plasma facing surfaces (b).
Details
regardingtheelemental
composition of plasma facing surfaces of thedivertor floor can be found in [6]. Only a briefsummary of the most important facts is given
here: (i) on many tiles the maximum inventory
exceeding 5x1019 D at cm-2 was detected in
narrow (approx. 5 mm wide) deposition belts inbeam was limited to approximately 8–8.5 µm
whereas thicker co-deposits with very highdeuterium content and the concentration ratio of
deuterium-to-carbon, D/C > 0.5, were detectedon some tiles.
Fig. 5 shows deuterium profiles on
poloidally and toroidally-oriented side surfaces.
The most important result is that the deposition
has occurred on the entire side surface. The
deuterium content varies in the gap but on
average it is not less than 3x1019 D cm-2. On thesame surfaces, the beryllium content issignificantly smaller: up to 3x1018 Be cm-2 in theregion 1-3 mm from the plasma facing edge andonly 0.3–1.1x1018 Be cm-2 in the lower part of thetile. Therefore, one concludes that the fuelretention on the surfaces in gaps is associatedwith the transport of hydrocarbons and formationof carbon-rich co-deposits. The analyses havebeen carried out for a limited number of tiles, but
the visual inspection of the whole poloidal set ofdivertor components has provided the evidence
that such a deposition pattern was typical. Basedon these results, the total integrated amount offuel on side surfaces is assessed at the level of
15.9x1023 D atoms, i.e. nearly twice greater thanon plasma facing surfaces.
III. A. 2. Mk-I Divertor with Beryllium Tiles
Images in Fig. 6 a-c show details of Be
tiles from the Mk-I divertor. Two features areimportant: (i) on the plasma facing surfaces thedeposition is particularly significant in the region
shadowed by the adjacent tile; (ii) deposition on
the side surfaces has occurred even at the bottom
part of divertor components
100

DEUTERIUM (e18/cm2)

the shadowed part of the tile (see Fig. 1); (ii) there
were no areas with the D concentration lower
than 5x1017 cm-2. Assuming toroidal symmetry ofthe deposition pattern, the total inventory on theentire area of plasma facing surfaces on thedivertor floor (11.1 m2, projected area of gaps
subtracted) could be assessed on the level of
8.86x1023 D atoms. This value should be treatedas a lower limit, because the information depth
for deuterium with NRA using a 2.5 MeV 3He+

80

Tile 1-tor
Tile 2-tor
Tile 2-pol

60

40

20

0
0

10

20

30

40

DISTANCE FROM PLASMA (mm)50

Fig.5. Distribution of deuterium on sidesurfaces in the gaps between the Mk-I CFC tiles.

Fig. 6. Castellated Be tiles from Mk-I: plasma
facing surfaces (a) and (b), side surfaces (c).
As previously reported [6], the greatest
deuterium content on plasma facing surfaces wasfound on tiles from the inner divertor corner:

from 3x1018 cm-2 in the non-shaded part to over
3x1019 cm-2 in the region shaded by the adjacenttile. Recent examination of these inner cornertiles has permitted, for the first time ever, theassessment of material transport to the sidesurfaces and castellation grooves in beryllium. In
order to enable the studies a few teeth ofcastellation were cut off. This allowed forcomparison of the fuel content on two kinds ofsurfaces: located in the groove and in the gap
between the tiles. Fig. 7 shows a detailed imageof the examined tile.

2x10196x10186x10178x1016

Fig. 7. Deuterium on side surfaces and incastellated grooves on the Mk-I beryllium tile.
The most important result is that
deuterium has always been detected together withcarbon indicating that its co-deposition occurredin the gaps and in castellated grooves. The shapeof the deposition profile in the gap area isqualitatively similar to that on CFC tiles (see Fig.
6), but the total D content is distinctly smaller
than on CFC tiles. This may be attributed to two
factors: (i) a shorter operation time of Mk-I with
Be tiles than with CFC components; (ii) a lack of
the local carbon source in case of the Be divertor
floor. The latter means that the carbon found in
the divertor originated from the erosion of themain chamber wall. In the narrow castellated
grooves the fuel content decreases by roughly anorder of magnitude over a distance of 5 mm, from
6x1017 cm-2 near the plasma facing edge to 8x1016
cm-2 in the region deeper in the groove. These

concentrations are 70-100 times lower than thoseon plasma facing surfaces [6] and in the gaps
separating tiles.
From the analysis data in [6] theintegrated fuel content on the plasma facing sideof the inner corner Be tile is estimated to be
9.32x1020 D atoms (Csurface). As recently
determined, the integrated content in thecastellation grooves of this tile is around0.224x1019 D atoms (Cgroove), whereas 3.05x1020
D atoms (Cgap) are found on the surface located in
the gap. One concludes that: (i) Cgroove/Csurface =0.024 (2.4%) and (ii) Cgap/Csurface = 0.33 (33%).
The results clearly indicate that the fuel content
in the shadowed parts of the beryllium divertor is
distinctly lower than in the CFC structure.
Secondly, the retention in the castellated grooves
is small in comparison to that on plasma facing
surfaces and in gaps between the tiles.
III. B. Deposition in the Mk-IIGB Divertor
III. B. 1. Vertical and Horizontal Tiles
Plots in Fig. 8 a-c show deposition
profiles of deuterium and beryllium in the innerleg of the Mk-IIGB divertor, i.e. on Tiles 1, 3 and4. One observes significant variations in the
distribution of species. On Tiles 1 and 3 it isrelated to the flux deposition profiles. The fuelcontent is in the range from 3x1018 to 1.8x1019 D
atoms cm-2. The situation changes drastically on
Tile 4 where two distinct regions are clearly
distinguished: (i) a very thick deposit in the area
shadowed by Tile 3 and (ii) a moderate level on
the part open to the plasma. As previously
reported [7], the deposit thickness in the
shadowed region reaches approximately 60
microns, whereas the results in Fig. 2 c reflect theconcentration in a layer 8-8.5 µm thick, i.e. thedepth accessible by IBA. Therefore, the actual
total content of D in that region is at least 6 times
greater and it amounts to about 7.2 x 1020 at cm-2.
Moreover, lack of Be in the shadow indicates thatthe deposit contains thick carbon film resulting
from a long-range transport of hydrocarbons.
Such a deposition pattern was identified anddiscussed following the studies of the Mk-IIA
divertor [4,5].
15

Be
D
10

Assuming a toroidal symmetry in deposition, theintegrated amount of retained fuel on all tiles
would be around 4.68x1024 D atoms including as
much as 3.63x1024 D atoms trapped in the
shadowed region on Tile 4. These numbers,
however, should be treated as a lower limit ofretention, because for technical reasons, thematerial deposition and fuel retention in someareas of the shadowed region in the inner divertor
could not be studied.

5

III.B.2. Deposition in the Septum
0
0

50

100

20

150

200

250

Be
D

15

10

5

0
0

30

60

90

120

150

180

100
Be
D

80

60

40

20

0

0

40

80

120

160

Fig. 8. Poloidal distribution of deuterium (opensquares) and beryllium (filled diamonds) on Tiles 1, 3
and 4 in the inner leg of Mk-IIGB.

Deposition and fuel retention on plates
from the outer divertor (6, 7 and 8) has beensignificantly lower than in the inner leg. The only
exception was a narrow belt of strong deposition
on Tile 6 in the region shadowed by tile 7.

Surface morphology of all verticalsupport and divider plates of one septum module
was examined. Fig. 9 a shows the appearance ofthe support plate (C-end, see Fig. 3 b). A dusty
deposit with high fuel content (25x1018 D atoms
cm-2) is found only in a narrow belt, 10-15 mm
wide, at the plate edge facing towards the innerdivertor channel. Similar deposition pattern has
been observed on other support plates. In somecases erosion by arcing has also been noted. No
significant deposition zone is formed on the outer
divertor side, where a constant level below
1x1018 D atoms cm-2 is measured. The Be content
across the plate is in the range 0.5-1.5x1018 atoms
cm-2 and it is not increased at the edge facing
towards the inner divertor. This indicates that the
thick co-deposit is mainly composed of ahydrogenated carbon film. There are some small
variations in the D and Be levels when all theanalysed plates are compared, but the general
distribution pattern remains unchanged.
Fig. 10 shows a divider plate and theamount of deposited species. These results are
representative for all the plates under
examination, both for the inner and the outer
divertor side. Again, the amounts of D and Be aresmall and the coverage is fairly uniform over that
surface: 0.5-1.5x1018 atoms cm-2. The analyses ofthe support and divider plates have given highly
consistent results showing low level ofco-deposition and, as a consequence, small fuelretention.

25

C O N C E N T R A T IO N [ 1 0 e 1 8 /c m 2 ]2

20

CONCENTRATION [10e18/cm2]

Be
D

Be
D

1.5

15

10

1

0.5

5

0
0

20

40

60

80

100

120

"TOROIDAL" POSITION [mm]

0
0

30

60

90

120

POLOIDAL POSITION [mm]

150

180

Fig. 9. A poloidal support plate of the gas box
module (a) and distributions of D and Be on theplate (b).
The total surface area of the gas box
components is 14.85 m2. Taking into account the
average D content of 1.5x1022 at cm-2 outside thedeposition belts and 12x1018 at cm-2 in the belts
and assuming the toroidal symmetry ofdeposition in the septum, the extrapolated upper
limit of fuel retention would be 3.28x1023 atoms.
When absolute numbers are compared, thisamount is over one order of magnitude smaller
than that assessed in the shadowed part of Tile 4,i.e. 36.3x1023 D atoms retained on an area ofaround 0.5 m2.

Fig. 10. A divider plate of the gas box (a) anddistributions of D and Be on the plate (b).
III.C. Comparison of Deposition in Mk-IIA
and Mk-IIGB Divertors
Mk-IIA and Mk-IIGB divertors were in
operation for a similar period of time. The
majority of retained fuel was found in the innerdivertors, especially in the zone shadowed fromthe direct plasma impact. Fig. 11 shows thedeposition histogram on the plasma facing
surfaces. A brief comparison of the deposition
allows for the statement that in both cases greater
deposition has occurred in the inner than in theouter divertor. The exception is a high deposition
region on Tile 6 being in the shadow of Tile 7 in
Mk-IIGB. However, there is a significant
difference in the deuterium-to-beryllium

concentration ratio when the composition of thesurface layer is compared: CD/CBe > 1 in
Mk-IIGB and CD/CBe< 1 in Mk-IIA. The
difference can be primarily attributed to the end
phases in operation with the divertors.
Immediately after the D-T experiment in JET
with Mk-IIA, the vessel was intensively
cleaned-up by various means in order to reduce
the tritium retention [1]. This certainly reduced
also the deuterium content in the surface layer.
On the contrary, no special cleaning procedure
was done after the operation with Mk-IIGB
because discharges were fuelled only with
deuterium and helium [8].
30
Be
D

20

10

0

30
D
Be

20

10

0

1

Fig. 11. Comparison of deposition on verticaland horizontal tiles of the Mk-IIA and Mk-IIGB
divertors. Results for shadowed regions on Tile 4
are not included.
IV. Concluding Remarks
An important contribution of this work
to tritium retention studies is the assessment offuel accumulation in structures similar to thoseforeseen in ITER: castellation of PFC and dome.
They have large areas shadowed from the direct
plasma impact and, potentially, can trap a lot of
co-deposited fuel. Until recently, the only
experience with large-scale structures of this kindhas been at JET operated with castellated tiles in
the Mk-I divertor and with the gas box module in
Mk-IIGB.
The essential conclusion derived fromthe comparison of deposition in the CFC and Be
Mk-I divertors is that fuel inventory in gaps andcastellation grooves is associated with thedeposition of carbon. Much more pronounced
accumulation has occurred in the presence of the
CFC structure than in case of the Be tiles. It is notintended here to give predictions for ITER, but
one may suggest that in a machine with
non-carbon walls in the main chamber thematerial transport and consequential fuelinventory in the castellation would be reduced.
The difference in fuel content in the gaps
between the tiles (less in Be than in CFC
structure) also indicate that the co-deposition in
the divertor is attributed to two factors: (i)transport of species eroded from the mainchamber wall; (ii) transport from the local carbon
source in the divertor itself. Small inventory inthe castellated grooves of beryllium tiles points tothe probable role of the gap width on the in-vessel
fuel retention. This issue will be the subject of
future studies.
The assessment of the deposition andretention inside the gas box was particularly
important because many diagnostics in ITER
plan to gain access through the septum structureand excessive deposition would obscure viewing
capability of optical systems. The measurements
clearly indicate that the septum region should not
be considered as a major trap for transported andco-deposited material. Fuel inventory in the
septum is insignificant in comparison to that in
the shadowed areas in the inner divertor corner.
From the very small deposition found on thedivider plates, one also concludes that the directcross-divertor transport of material (neutral
atoms and molecules) through the dome, from theouter to the inner leg, can be considerednegligible.

These results from JET operated withthe carbon wall and evaporated beryllium coatingin the main chamber should not be immediatelytranslated into conclusions and predictions
regarding the material migration and fuelinventory in the ITER divertor. However, theresults allow for some optimism showing that no
dramatic deposition inside the dome structure
would occur, especially, since the use of carbon
in the next-step machine is planned to be strongly
limited. According to present materials choice forin-vessel components of ITER [9], the source ofcarbon will be significantly reduced (bulk Be
tiles on the main chamber wall). This will change
the scenario of material erosion and will also
influence the co-deposition of fuel.
Acknowledgements
This work has been conducted under
the European Fusion Development Agreement
and is partly funded by EURATOM, the UK
Enginnering and Physical Sciences ResearchCouncil and the Swedish Research Council (VR)under the Contract VR-621-2003-3170.
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