Tritium Retention in Plasma Facing Graphite Tiles
T. Tanabe1, K. Miyasaka2, T. Shibahara2, R. Ishida2, N. Miya3, K. Masaki3, V. Philipps4 and M. Rubel5
1

CIRSE, Nagoya University
Graduate school of Eng. Nagoya Univ.
3
JAERI, Naka
4
Juelich Research Center, Germany
5
Van Alfven Laboratory, Sweden

2

Abstract
The present work, at first, summarizes the basic
knowledge of hydrogen behavior in graphite and thenintroduces recent investigations how tritium is
distributed in plasma facing walls of TEXTOR and
JT-60U in terms of material temperature (or incidentflux), local plasma behavior, global plasma behavior
and so on.
In TEXTOR, it was found that tritium
distributions was quite different from deuterium
distribution, i.e. deuterium retention was higher at thedeposited area, whereas tritium retention was higher atthe erosion dominated area. This is because tritium
produced by the D-D reaction, initially having 1 MeV,
did not fully lose its energy in the TEXTOR plasma
and implanted into the plasma facing materials nearlyhomogeneously, whereas deuterium was codeposited
with carbon and boron, the main impurities in the
TEXTOR plasma. This is also confirmed by thefinding that high level of tritium was detected beneaththe deposited layer.
In JT-60U, tritium distribution, however, wasmodified by the temperature increase due to plasma
heat load. The highest tritium level was found at the
top of the dome or the private region and the outerbaffle plates, where the plasma did not hit but the
distance from the plasma was the shortest. For thedivertor tiles, the tritium retention was very small.
Such tritium distribution observed in JT-60U tiles canbe well explained by the homogeneous implantationof rather high energy tritium and thermal release dueto the heat load.
Thus the comparison of tritium profiles with thedeuterium profile gives a large amount of importantand new information on PMI, and may be used as a
new diagnostic technique for PMI.

I. Introduction
Tritium retention in plasma facing materials bothfor long term and short term is one of the mostimportant safety issues in fusion reactors [1]. After theuse of tritium in JET and TFTR, the immediate
retention was about 40% and 51%, respectively [2, 3].
Even after the excessive tritium removal procedure
applied to both machines (D-operation, D2 andHe-glow, O2-venting), remaining long-term tritium
retention of ~16% and ~13% was observed,
respectively. A serious concern is the very hightritium retention in deposited layers at lower
temperature region, particularly in the deposited layers
in divertors and louvers in JET [4]With the use of a scintillation method (whichneeds sophisticated procedure, full combustion oroxidation, accumulation of tritiated water vapor andscintillation measurements), a large set of wall tiles
have been analyzed and tritium surface distributions
and depth distributions in the plasma facing tiles havebeen reported [4,5]. Unfortunately, this technique can
not give detailed surface profiles and depth profiles oftritium, which is very important to the understandingof tritium retention and transport in tokamaks.
Very recently we have succeeded in making
tritium images on graphite tiles used in TEXTOR[5,6]and JT-60U [7] by means of an imaging platetechnique. In both machines, tritium is produced by
the D-D reaction. Hence energetic triton without
fully losing energy was implanted to all of plasma
facing walls, but the tritium profiles in the wall was
clearly modified by the plasma conditions andcharacters, and the wall temperatures as well.
The present work, at first, summarizes the basic
knowledge of hydrogen behavior in graphite and thenintroduces recent investigations how tritium is
distributed in plasma facing walls in terms of material
temperature (or incident flux), local plasma behavior,global plasma behavior and so on. Comparison with

deuterium distribution determined by ion beam
analysis is also mentioned to give very important
information on the behavior of tritium.

II.

Carbon-Hydrogen ChemistryFig.1 shows schematic of hydrogen retention
on plasma facing tiles[8]. Because of porous natureof graphite, energetic hydrogen and molecularhydrogen behave quite differently, i.e. energetichydrogen impinges surface and stops at aroundprojected range, whereas molecular hydrogen canpenetrate deep inside along open pores. As a result
gaseous hydrogen can permeate through graphite.
However, very small diffusivity keeps hydrogen
only near the surface region of the grains.
Therefore depth profiles in large scale can be
written smooth change as show in Fig.1, even thedetailed- or or micro-scale profiles in each grains
are significantly different.
In addition hydrogen retention in graphite isquite temperature dependent. Below 500K implantedhydrogen cannot move to be trapped and the surface
region is easily saturated with hydrogen concentrationof about 0.4 for H/C, while above 800K hydrogen
diffusion becomes dominant and most of implantedhydrogen is re-emitted remaining very small amountof hydrogen.
It is also noted that once hydrogen molecules areproduced on the grain surface through a mechanismthat migrating atomic hydrogen recombines withtrapped hydrogen, the hydrogen molecules canpenetrate to the top surface to be released out.
Between 500K – 800K hydrogen retention quite
depends on temperature. Accordingly C-H filmshows very wide range of H/C ratio as shown in Fig.2,
where carbon is characterized in to two different

species, i. e. sp2 and sp3.[9] Still H/C ration in
production of C/H films is not free, but limited incertain range depending on the producing methods
and temperature.
Below 500K, graphite irradiated with hydrogenup to saturation level is reported to show the samecharacter as that of hydrogen saturated amorphous
carbon (a:C-H) film.[10]With increasing thesubstrate temperature, H/C in the deposited film
decrease, and becomes pyrolytic carbon, well but not
fully graphitized carbon.
It should be noted that in JET and JT-60U thefirst wall is operate below 600K while TFTR around
400K or below. This makes tritium profile ofin-vessel component of these machines a little
different as discussed later.

Ⅲ . Tritium detection by imaging platetechnique
The imaging plate (IP) used here was
BAS-TR2025 for low energy β-rays emitter such astritium, manufactured by Fuji Photo Film Co. Ltd.
The surface of the IP was contacted to the graphite tile
surface for one hour in a dark shielded room. For
full toroidal measurements for the tiles of the top of
the dome units, we have inserted a thin (2 μ m)
poly-film between the tile and the IP, which reduces
the sensitivity by a factor of 10. After the exposure,the IP was processed by an imaging plate reader, Fuji
BAS-2000 or BAS-2500 to obtain digitised intensity
mapping (“tritium image”) The characteristics ofimaging technique and detailed measuring methods
are given elsewhere. [7,11]
Ⅲ-1. Tritium retention in plasma facing graphite tilesof TEXTOR
Ⅲ-1-1 Tiles measured
Graphite tiles investigated here were taken fromtwo different types of limiters used in TEXTOR-94:(ⅰ) two graphite tiles (tile No. 8 and 22 at blade 4) at
the middle position of the forth blade of the toroidal
ALT-II limiter and (ⅱ) two (the top and the bottom)
tiles from the bumper limiters. Two collector probes
made of Si at the top and the bottom of the torous,
which were not directly exposed to the plasma but
faced to the plasma directly, are also examined. The
locations and the configurations of the limiter tiles areshown in Fig.3.
All tiles (from both the ALT-IIlimiters and bumper limiters) were used in theoperations from the middle of 1998 to the middle of1999 and exposed to more than ten thousands shots.
Tritium detected was within the depth of 3.5μm fromthe top surface of the graphite tiles and a little less in
the Si collector probes. The details of the tritium
imaging plate technique (TIPT) were given
elsewhere[6,7] and also in this conference.[11]Deuterium distribution was measured by means of anion beam technique, as given in Ref.[12].
Ⅲ-1-2. ALT-II limiter tiles
Fig. 4(a) shows the tritium image for one of the
ALT-II limiter tiles (No. 22). Vertical and horizontal
directions correspond to the poloidal and toroidal
direction in the torus respectively. Two holes in thecentral area are for the bolts fixing the tile to thebase plate. The lower, larger part of the tile is theerosion-dominated region and the upper part with arough surface is the deposition-dominated region
covered with redeposited layers with a thickness ofFig. 3. Photograph of TEXTOR-94 vacuum vessel
with ALT-II toroidal belt limiters, bumper limiters.
Two tiles of No.8 and 22 at blade 4 of ALT-II limiters,two (the top and the bottom) of the bumper limiters areshown in the figure.

around 5 µm [11]. As seen in the line profiles,
tritium distribution at the erosion-dominated region isnearly homogeneous and much higher than that at thedeposition-dominated region. However, by peeling offsome area of the deposited layer (in the upper left
region of the tritium image), the tritium level beneaththe deposited layer was again homogeneously
distributed with tritium levels a little higher than thatof the erosion-dominated region as seen in the line
profiles.
It is interesting to compare the tritiumprofiles with deuterium and boron profiles given by
the ion beam analysis of one of the ALT-II tiles. (SeeFig.4(b)) The tile analyzed by the ion beam was not
the same one as used for tritium imaging, but there
were no significant differences in these profiles
among the different ALT-II tiles. As clearly seen in
Fig.4(b), deuterium and boron show quite similar
profiles and their densities are higher at theredeposition dominated region (poloidal position >〜 80 mm). In another word, deuterium wasretained in a very shallow region by re- or
co-deposition with boron and carbon, whereas tritium
was retained deeper.
Tritium was also detected even at the backside of
the tiles where was not directly attached to the baseplate (not shown here). Tritium distribution was still
homogeneous, but the amount is an order ofmagnitude less than that for the plasma facing side.
The results for another tile (No. 8) were quite similar
except that a little higher tritium concentration wasobserved (not shown here).
Ⅲ-1-3 . Bumper Limiters
Fig. 5 shows the tritium images of the top and thebottom tiles from 5 bumper limiter tiles together withthe toroidal line profiles of tritium and deuterium forthe bottom tile (Fig.5(d)). Because of the saddle type
shape of the tile, the image was divided into theelectron drift side and the ion drift side. Although thetritium level of the electron drift side is a little higher
than that of the ion drift side, the tritium distributions
within either side were both nearly homogeneous.
One can clearly see that the tritium level issignificantly higher for the bottom limiter than for the
top.
Tritium was also detected even at the backsideof the bumper limiters.
As clearly seen in the Fig.5(b), the profiles fortritium and deuterium were completely different.

The appearance of the bumper limiters did not show aclear deposition pattern, but there might be some
deposition, as indicated by the observed high retentionof deuterium. The temperature of the bumper limiter,
which was not measured, also likely influenced the
profile as will be discussed in the next section.
Ⅲ-1-4. DISCUSSION
A. Homogeneous distributions on the ALT-II tiles
In the previous work [6], we have suggested that
tritium produced by D-D reaction, initially having theenergy of about 1 MeV, did not fully lose its energy in

Deposited region

BeneaththetheBeneathdeposited layer

Radioactivity [PSL/mm2]

ErodedErodedregion
region

300

0
100
200
300
Radioactivity [PSL/mm2]

200
100
0
0

50

100
Scale [mm]

150

(a)

(b)

(a)

(b)

(c)

10
8

D

T T

D

100

6

50

4
2
0
0

100
50
Distance from "LEFT" edge [mm]0T radioactivity [PSL/mm2]

D content [1017 atoms/cm2]

Fig.4. (a) Tritium image for a graphite tile (No.22 at blade 4 of the ALT-II limiter), together with line profiles along thepoloidal (vertical) and toroidal (holizontal) directions. (b) Deuterium and boron line profiles along the poloidal direction
together with the cross section of the tile.

(d)Fig.5 (a) Photograph of the top and bottom of the bumper limiter tiles, (b) tritium images from the front surface and (c)from the back side of the respective tiles. (d) tritium (−) and deuterium (Δ) line profiles along the toroidal direction
at the front surface .

the TEXTOR plasma (with a temperature of about
2keV, density 3 x1020 /m3, and the magnetic field of
2.2 T), and implanted into the plasma facing materials
nearly homogeneously. This was confirmed by theobservation that the high level of tritium retention wasobserved behind the deposited layer. That is because
the triton (tritium ions) gyrates in the magnetic fieldwith a maximum gyro-radius of about 11cm, and canpenetrate into the gap or turn around to the back sideof the limiters where it was not directly facing toplasma but had some space between the tiles and theback plate of the liner.
It was rather surprising to find that the tritium
distribution was quite different from the deuterium
distribution [11] (See Fig.4(b) and Fig.5(d)).
However deuterium analysis clearly showed that
deuterium was codeposited with carbon and boron, themain impurities of the TEXTOR plasma. This is
another confirmation that tritium produced by the D-D
discharge behaves differently from those tritium
introduced as a gas fuel or by NBI.
B. Asymmetry along poloidal direction and thedifference between ALT-II tiles and bumper limiters
As discussed above, tritium distribution can begenerally described by the homogenous energetic
implantation from the plasma but not codeposition nor
redeposition.
Nevertheless, the poloidal tritium
distribution obviously shows deviations from thegeneral tendency. The tritium intensity of the bottom
bumper limiter was clearly higher than that of the top
one. Such asymmetry was more clearly shown in thecomparison of the collector probes. This asymmetry
should be attributed to the higher density of the lower
half of the plasma than the top half, though there
might be some contribution from the gyration and the
direction of the magnetic field.
Due to the highpenetration power of neutrons, it might be difficult
to use neutron profiling to see such asymmetry,Thus, tritium profiling for the full poloidal direction
could be a good diagnostic technique for plasma
profiling in D-D discharges.

bumper limiter could be subjected to a little higher
temperature than the ALT-II limiters. During NBI
discharges the heat load of the ion drift side is likely
higher than the electron drift side and the saddle type
shape of the bumper limiter can separate the heat load.
Thus tritium retention on the bumper limiters wasclearly different for the electron drift side and the ion
drift side.
Such a profile change due to the temperature rise
was also seen in tritium retention at the leading edgeof the ALT-II limiters. One can see that the tritiumprofiles decay more slowly at the leading edge
compared to the steep decay at the side edges. (seeFig.5) The reason for the tritium retention at theNo.8 graphite plate being a little higher than that at
the No.22 graphite plate is probably owing to a higher
heat flux to the No.22 plate than the No. 18 graphite
plate [13], resulting in a little higher temperature andless tritium retention for the No.22 plate.
On the bumper limiters, it is difficult todistinguish the erosion dominated regions form theredeposition dominated regions. Analogous to the
ALT-II limiter, those regions with higher deuterium
retention were probably dominated by redeposition.
At high temperatures, deposition results in slightly
graphitized carbon with good adhesion to the substrategraphite. According to the ion beam analysis, theD/C ratio at the low deuterium retention regions in the

C. Target temperature effect
One can note that deuterium retention on theredeposited region of the bumper limiters is one orderof magnitude less than that of the ALT-II limiters(compare Figs. 4(b) and 5(d)).
Although there
might have been some difference in the incident
deuterium fluxes between the two, such a large
difference in deuterium retention could not havebeen due to the flux difference entirely.
In fact,
the tritium retention, which should be nearlyhomogeneous, was also higher for the ALT-II limiters
than for the bumper limiters. This indicates that theFig.6 Configuration of graphite tiles in JT-60U divertor
regionbumper limiters was 〜 0.05 [14] (see Fig.5(d)),
indicating that the temperature should be above 900K.
In toroidal direction, particularly near the rightedge, the tritium retention was slightly less, which
could also be due to the higher heat load. However,
the tritium retention was still very small in the tiles, i.e.
far below the saturation, which means that without
significant temperature increase it would be difficult
to discuss the temperature effect.
In a separate
work of tritium analysis of the divertor tiles of JT-60U,
such temperature effect was more clearly observed[7,15].

respectively. However, surface temperature during
plasma discharge was not measured for all tiles.
The tiles were exposed to deuterium plasma inthe period of June 1997 to October 1998. During
this period, about 1x1019 neutrons were generated by
D-D reactions. Accordingly 1x1019 or 18 MBq oftritium was also produced. The vacuum vessel of
JT-60U was subjected to hydrogen discharges toremove the tritium retained in the vacuum vessel, thenfollowed by air ventilation before fully opened to theatmosphere. Thus long term tritium retention in thevacuum vessel was about 50% of the total production.
[15]

Ⅲ-2 Tritium retention in plasma facing graphite tilesof JT-60U

Ⅲ-2-2. Results and Discussion

Ⅲ-2-1. Tiles measured
Graphite tiles examined here were isotropic
graphite (IG-430U) and CFC graphite (CX-2002U)
used at (1) divertor region taken from the poloidal
direction including the baffle plates and (2) the top of
the dome units taken from the toroidal direction.
The configuration of the tiles in JT-60U is given in
Fig.6.
These tiles were fixed to the base metal bybolts and their temperature was measured bythermocouples located at the backside.
The
temperature increase due to the plasma heat load wasaround 50 K except at the divertor striking points
where the maximum temperatures of 800 and 1200 K
were recorded at the inner and outer divertors,

Fig. 7 compares the photographs and tritium
images of the divertor units and the baffle plates. In
this figure, tritium level is higher in the red region andless in the blue region. The tritium levels at the topof the dome unit and baffle plate were estimated to bein the order of 10kBq/cm2.[7]Fig.8 shows corresponding tritium line profiles. It
was rather surprising to see that the highest tritium
level was recorded at the outer baffle plates and the
top of the divertor dome, where plasma did not
directly hit. The tritium level at the divertor tiles
were very small, particularly both the outer and inner
strike points showed the lowest level owing to thetemperature escalation during plasma discharges.
This will be discussed in the next section. The tritium

Fig.7 Photographs and tritium images of divertor units and baffle plateslevel at the top of the dome units is homogeneous,while the tiles of the both sides show steep gradient in
the poloidal direction.
Suppose the total produced tritium (18GBq) in
JT-60U was distributed homogeneously over theplasma facing area, the areal tritium level could be
estimated from the distance of the tritium source
plasma. However, not all the tiles face the plasma
directly and the distances from the plasma also dependon the geometrical configuration of the tiles in thetorus. Thus we can estimate tritium flux assumingthat the tritium is homogeneously produced in theplasma and injected to all tiles, taken into account the
distance and the angle of the incidence. The results
are given by solid lines in Fig. 8. Approximating
the D shape of the plasma by a circle over full torus
with a minor radius of 1.4 m and the major radius of
3.3 m, the total surface area is calculated to be 180 m2.
When totally generated tritium of 18 GBq is divided
by 180m2, areal tritium distribution becomes around
10 kBq/cm2. This is the same order of the magnitude
as the tritium level observed, confirming that largepart of the tritium produced by D-D reaction was
injected homogeneously over the plasma facing
surfaces without fully losing its initial energy of 〜
1MeV. The small differences among the tritium
retention in all of the tiles of the top of the dome
unites over 2π toroidal direction (in Fig.9) except
some local structure also confirms the homogeneous
implantation.

In Fig.8, therefore, the estimated tritium fluxes
were normalized at the highest tritium level.
Although the estimated tritium flux correspond to thegeneral trend of the tritium profiles along the poloidal
direction, the deviation particularly at the divetor tiles
is appreciable. One may attribute the deviation to
the temperature increase of the tiles due to the plasma
heat load. Unfortunately surface temperature of thetiles was not measured.
Then we have estimated the surface temperaturefrom the retained tritium level, to confirm the above
argument using following assumptions: ( ⅰ ) thetritium retention together with deuterium andhydrogen at the highest point corresponded to thesaturated hydrogen level in graphite [16] at 600K (themeasured temperature by the thermocouple at the backside of the tiles), and (ⅱ) tritium once implanted was
released by the temperature rise of the tiles due to theplasma heat load. We can estimate the surface
temperature based on the temperature dependence ofthe saturated hydrogen retention [7] and shown in
Fig.5, for the tile at the outer side of the dome units.
Once we had obtained the surface temperature(calculated and shown in Fig. 9) , we could also
calculate the heat flux based on the steady state heat
flow, with the backside temperature given by thethermocouple, and the thermal conductivity ofgraphite 〜80 Km-1s-1. The calculated heat fluxes
which are around 1 to 5 MWm-2 depending on theposition, are also plotted in Fig. 5. The calculated

Comparison of tritium retention and estimated tritium flux
250
Estimated Tritium flux

200
Baffle plate

Dome unit

Inner divertor

Outer divertor

Estimated Tritium flux /Tritium Retention / PSL mm-2

Tritium line Profile

Baffle plate

150
Normalized point

100

50

0
-600

-500

-400

-300

-200

-100

0

100

Horizontal Distance / mm
Fig.8 Tritium line profiles along the poloidal direction in the divertor region. Estimated tritium flux is also shown(see text). Blue and red lines are corresponding to two different line profiles on the same tiles.

Outer Dome Tile

In order to investigate tritium behavior intokamak, we have measured the surface distributionsof deuterium and tritium on graphite limiter tiles usedin TEXTOR under D-D operation, respectively, bymeans of an ion beam analysis and tritium imaging
plate technique.
Tritium produced by D-D reaction did not fully
lose their energy before implanted in the plasma
facing surfaces. As a results tritium distribution on
the tiles were generally homogeneous. High energy

Tritium Retention (Arb.Units)

SurfaceTemperature

600

100
500
Heat Load

80

400

60

300

Expected T retention at 300K

40

200

20

100

0
-250

4.0

Heat Load MW/㎡

IV. Conclusion

700
120

Temperature ℃

values are quite reasonable compared to the separate
estimation of the heat load [14]. The slightly higher
tritium level at the inner side edges in all tiles of the
top of the dome units (see Figs. 8 and 9) might be also
explained by the larger temperature increase observedat the outer divertor than that at the inner divertor [7].
Thus tritium distribution observed in JT-60U tiles
can be well explained by the homogeneous
implantation of rather high energy tritium andsimultaneous thermal release due to the heat load.
Accordingly, the tritium distribution at those regions
which are near the plasma but not directly in contact
with the plasma (such as the top of the divertor andbaffle plates) retained high level of tritium instead ofthe divertor tiles where the tritium was very small.
This is also confirmed by the nearly same tritium
images for all tiles of the top of the dome measured
here except some local structure in torus direction.
As shown above, in TEXTOR, tritium retention
in the redeposited layer was relatively small. On theother hand, most of the tritium images of the JT-60U
tiles showed no clear patterns but gradual changes,
whereas some patterns probably owing to the erosionand deposition were appreciable in the photograph ofthe tiles.
In the divertor tiles, because of the theirtemperature increase, properties of the deposited layer
may be somewhat different from the deposits at lower
temperature, or the thickness of the deposits might bemuch less. These must be investigated in future.
In this tritium imaging plate technique, it is very
likely that IP imaged missed some tritium adsorbed orabsorbed in the near surface layer from low energy
impinging after losing energy in the plasma.
This isbecause such surface tritium was easily replaced bydeuterium during the discharges and was mostly
removed by hydrogen discharges just before the torus
opening and the subsequent air ventilation. This might
be additional reason why the tritium distribution was
homogeneous without showing any appreciable
inhomogeneity in the edge plasma. Thus most of thetritium detected by IP was those retained rather stablein the graphite from the top surface to a depth ofaround 5 μm.[10]

2.0

0
-220

-190

-160

Distance/mm
Distance/mm
Fig.9 Comparison of tritium profiles, expected tritium
retention, estimated surface temperature and heat flux
from the tritium content (see text) for one of the outertiles of the dome units.

injection was confirmed by the observation of tritium
beneath the deposited layer. Tritium was detectedeven in the plasma shadow, probably owing to their
gyration in the scrape-off layer.
The profiles of
deuterium were completely different from the tritiumprofile. Deuterium are co- or re-deposited withcarbon and boron on plasma facing surfaces at thedeposition dominated regions, whereas little
deuterium was observed at the erosion dominatedregions.
Tritium distributions on the graphite tiles used as
plasma facing tiles in divertor tiles, dome units, andthe baffle plates of JT-60U were also successfullymeasured. The highest tritium level was found at the
top of the dome or the private region and the outerbaffle plates, where the plasma did not hit directly but
the distance from the plasma was the shortest. Suchtritium distribution observed in JT-60U tiles can beagain explained well by the homogeneous
implantation of rather high energy tritium and thermal
release due to the heat load.
Therefore tritium
distribution at those regions near the plasma but notdirectly in contact with it (such as the top of the dome
unites and the baffle plates) retained higher amount oftritium. The tritium retention in the divertor tiles
was very small, particularly the tritium level at the
strike points of the divertos were near the detection
limit. Different from TEXTOT, no clear patterncorresponding to the deposition was observed in
JT-60U divertor tiles.
The observed tritium distribution reflects thetemperature history of the PFM, and the retention
seems very small at those regions that haveexperienced temperatures over 800K. The tritium
distribution also indicates plasma asymmetry, i.e.

higher density in the bottom half than the top half ofthe plasma in TEXTOR and some local structure in
the full toroidal tritium profiles of JT-60U, which mustreflect plasma character.
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