Tritium Retention in Plasma Facing Materials
C. H. Wu
Aachen University, Germany
In the next step D/T fusion device, the accurate predictions of tritium retention inPlasma Facing Component is important both from the point of view of the safety and thephysics performance.
However, for reliable prediction of T-retention, a detailed knowledge andunderstanding on the mechanisms, which are responsible for T-retention is essential. The
mechanisms are quite complex, for example, in fusion devices; the PFC is in contact withplasma and PFC surface exposed to large flux of ion and neutral particles. The large fluxof ions may be retained in the PFC materials by implantation in the depth of ion range,
the particles will diffuse to the bulk of materials and eventually, the particles are trapped.
In addition, the ion and neutral particles cause the sputtering of surface. The sputteredparticles are ionized in plasma and redeposited somewhere on the surface within thefusion devices. The sputtering and deposition lead to T-retention via co-deposition.
After a decade experimental investigation and theoretical study on D/T plasma wallinteraction, D/T transport in materials as well as neutron effects, the progress onunderstanding of the complex of T-retention is significant, so that the prediction of Tretention can be performed more accurate.
In this paper, a detailed analysis on the mechanisms of tritium retention has beenperformed, based on:
1. Tritium retention via implantation,
2. Tritium retention via bulk diffusion,
3. Tritium retention via neutron transmutation,
4. Tritium retention via co-deposition,
5. Tritium retention via neutron damage induced trapping sites
6. Tritium retention in dust and flakes,
Keywords: tritium, plasma facing materials, implantation, co-deposition, neutron effects
I. Introduction
The characteristics of tritium retention
and release of plasma facing materials are
extremely important from an environmental,
safety, economical and plasma density control
point of view.
As next step fusion devices, i.e. theInternational Thermonuclear Experimental
Reactor (ITER), unlike present tokamaks
devices, will be long pulse, high heat flux,
higher PFC temperature and they will produce
intensive neutron fluxes. Substantial R & D isstill needed to elucidate the influences of longpulse, high heat flux, high target temperatureand intense neutron on PFC tritium retention
and eventually on the total tritium in thedevices.
The present fusion devices with shortpulse (< 20 second), and low target
temperature, the tritium retention will bedominated by implantation, whilst the bulk
diffusion will be the major mechanism oftritium inventory for next step devices withlong pulse ~1000 seconds and higher target
temperature >350 ˚C. The bulk retention willcontribute significant to total tritium inventory.
Erosion and redeposition of PFC during
normal and off-normal operation caused
tritium retention via co-deposition of tritium inplasma and redeposited particles, which basedon present knowledge, will play also dominantrole.
Intense neutron will induce damage on the
microstructure and critical properties of PFC
materials[1,2] e.g. thermal conductivity,swelling and consequently the behavior oftritium trapping will be changed. Dependingon neutron damage, they could limit theutilization of PFC materials in the next step
fusion devices. The another issue is tritium
production by neutron reaction (transmutation)with PFC materials, which, depending on PFC
materials, may be significant.

The other important mechanism is tritium
retention in dust or flakes. During the normal
operation, the PFC will be sputtered by plasma
and re-deposition occurred and dust is formed
at PFC surface. Flakes are formed during theoff-normal events; e.g. disruption, ELMs or
run-away electrons by peel-off of layers oversaturated with hydrogen isotope.
In this paper, a detailed analysis on thepossible mechanisms of tritium retention has
been performed and relevance of results isdiscussed.
II. Results of analysis
II.A. Implantation
The tritium retention via implantation isspecial relevant to present tokamak device,because it is short pulse machine and low PFC
temperature. The most of particles are stopped
at the depth of ion range, which is dependent on
impinging particle energy. After theimplantation there is almost no further diffusion
to the bulk. At the relevant energy for next step
PFC (≤ 200 eV), the ion range D into beryllium
or carbon is around 5-10 nm. The saturationlevel of retention is a function of energy andtarget temperature. Database is very wellestablished. Fig. 1 and Fig. 2 show the ion
range of D into Be and carbon as a function ofion energy[3]. It is seen that the two low Z
materials behave very similar. The total amountof retention via implantation is proportional tothe fluency and impinging particle energy.

Fig.1 Mean ion ranges of deuterium in beryllium

II.B. Retention via bulk diffusion
In the next step fusion devices, theoperation conditions will be significantly
different from the present tokamaks: long pulse~500-1000s, or steady state operation, higher
target temperature (because of higher heatloading ≥ 10 MW/m2). Therefore, the role ofbulk diffusion will make an important
contribution to total tritium retention. The
diffusivity D (cm2/s) is expressed as : D = D0
exp (-Ed/RT), where D0 is pre-exponential
factors and Ed is the activation energy. The total
amount of retention via bulk diffusion iscorrelated to the solubility of materials which isexpressed as S = Ks P1/2 where Ks is Sieverts
constant and Ks = Ks0 exp (-Es/RT). Ks0 (at. fr.
Pa-1/2) is a pre-exponential factor, Es isFig.2 Mean ion ranges of deuterium in carbon

wall via transmutation can be as much as 1.5 kg(for the case: Be first wall: 1000 m2 area, 1 cmthick)heat of solution. In general, the lower thetemperature, the higher the retention via
solution. However, at low temperature, theexpected diffusivity is low, the time needed
to reach the saturated values is longer. At atarget temperature lower than 1600 K, thebulk saturation has not been reached eventhe exposure time as long as 108 seconds,
because the low diffusivity of H-isotope incarbon based materials. Whilst thesaturation value of retention in beryllium
has been reached even after only 103
seconds, because of the high diffusivity.
II.C. Tritiumtransmutation

retention

via

II.D. Tritium retention via co-deposition
During the normal operation in fusion
devices, the surface of wall materials will besputtered constantly by bombarding withplasma (mainly D, T ions). During thedeposition of the sputtered particles of C, Be
will be co-implanted with plasma, which leads
to tritium retention. Depending on target
temperature or oxygen concentration in plasma(particular for beryllium) [4,5,6]. The retention
can be as high as D/C= 0.4 and D/Be = 0.37 at
300 K. Fig. 3 show the D/T retention via codeposition as a function of target temperatureand oxygen concentration. Most recent resultshows, that there is no deposition (codeposition) by sputtering of carbon by plasma,
at target temperature ≥ 100 C.

neutronIn the next generation fusion devices, i.e.
the International Thermonuclear Experimental
Reactor (ITER) will produce intense neutron
fluxes via D-T reaction, which react with wallmaterials and may produce tritium, depending
on concentration and target temperature, tritium
will retain in the bulk of wall materials. For
example, at present design of ITER, Be will be
the first wall material. Several nuclear reactions
contribute the tritium production, anddepending on reaction time, the role of eachreaction is significant different. In the first 24hours, the 9Be(n,t)7Li reaction contributes morethan 99.85% of total tritium production, thereaction 9Be(nα)6He(β)6Li(n,α)t is dominant
after 3 years. The contributions of all reactions
as a function of reaction time are given in Table
1. The total tritium production via
transmutation in ITER Be first wall as afunction of reaction time is given in Table 2. It
is seen, that the tritium production in Be firstFig. 3 D/T retention via co-deposition

Reaction TimeReaction
24 h

1a2a3a5a

99,85

66,87

52,57

44,60

35,92

Be(n,∝) He(ß) Li(n,∝)t-

31,95

44,88

50,30

54,77

9-0,881,96

2,84,38

97Be(n,t) Li

966

3Be(n,X)t(ß) He(n,p)t

Table 1 Tritium Production via Neutron Transmutation (%)

Tritium (g)

24 h
0.675

Reaction Time
71a115.74 d (1.10 Sec)90.624360.838

3a1544.440

Table 2 Tritium Production in The ITER Be First Wall ( 1000m2, 1 cm thick)Material

Loading atmosphere

Loaded Tritium
Notirradiatedirradiated6

H2 + 5 ppm T2, 2 bar,850 ºC, 6-7 h(Bq/g)(T/Be)~ 1 x 10-8
~ 0.8 x 10T2, 2 bar, 850 ºC, 6-7 h(T/Be)~ 1.6 x 10

T2, 0.8 bar, 850 ºC, 10 h(T/C)~ 1 x 10(100 appm)

7~ 3 x 10-7~ 2.5 x 10

Be-3-4Graphite-2~ 5 x 10(~ 10000 appm)-3~ 1 x 10(~ 1000 appm)

Table 3 Neutron induced tritium trap sites in carbon and beryllium
II.E. Tritium retention via neutron damage
induced trapping sites

In several studies on the effects ofneutron induced damage on the tritium
retention have shown, that the neutron
damage on wall materials (C, Be) will
increase the tritium trapping sites. Thetritium retention increases with increasingof neutron damages. Fig. 4 shows thetritium retention as a function of neutron
damage and irradiation temperature. It isseen, almost independent on irradiation
temperature, at a damage of 0.5 dpa, asaturation value of 1000 appm has beenreached for carbon, whilst a saturation
value of beryllium is 10000 appm. The
comparison on the neutron damage
induced tritium trapping sites in C and Be
is given in Table 3. It is clear, under sameneutron irradiation conditions, the neutron
damage induced tritium trapping sites inBe is almost one order of magnitude higher
than that of in carbon [7,8].

II.F. Tritium retention in dust and flakes

Dust and flakes are formed either by
sputtering at normal operation or at offnormal operation, i. g. disruption, runaway electron events. The results of asystematic study show, that the tritium
retention in dust decreases with increasingparticle size, and also decreases withincreasing temperature. The tritium
retention in dust as a function of particle
size and temperature is given in Table 4.
The results of detailed study on dust
formation in JET showed, that average
particle size of dust is about 27 µm [9,10].
Tritium pressure ≈ 3.5 x 10-3 Pa
Exposure time ≈ 1.0 x 105 sSize(µm)

BET
2 -1(m g )25 °C

600 °C

800 °C

4.5

36.7

1.54

6.11

3.22

20

15.9

0.45

1.7

0.83

80

11.1

0.35

1.6

1.1

Table 4 Experimental investigation of Tretention in dust as a function of particle size(BET) and temperature (Tritium retention incarbon dust mg/kgC)

Fig. 4 T-retention as a function of neutron
damageIII. Conclusions

References

In this paper, a detailed analysis on themechanisms of tritium retention in plasma
facing materials has been performed.
The tritium retention via implantation is adominant mechanism for present tokamak
device, because the low wall temperature andshort pulse operation. In addition, the D/Tretention in flakes may be important, depending
on plasma condition. However, as next
generation devices, i.e., the International
Thermonuclear Experimental Reactor (ITER)
will produce intense neutron fluxes, thetemperature of plasma facing components ishigh, long pulse or steady state operation, themechanisms of retention fusion devices are acomplex. The bulk diffusion, neutron damage
induced trapping sites, co-depositions will play
dominant role and contribute major fraction tototal tritium inventory.

[1] J. P. Bonal and C. H. Wu, J. Nucl. Mater.
228,155(1996)[2] C. H. Wu, J. P. Bonal, H. Werle, FusionEng. Design,49-50,383(2000)[3] M. Mayer, W. Eckstein, Nucl. Inst. andMeth.B94, 22 (1994)[4] W. Wang, W. Jacob, J. Roth, J. Nucl.
Mater. 245, 66(1997)[5] C. H. Wu, A. A. Haasz, W. Jacob, R.
Moormann, V. Philipps, J. Roth, Recent
results of EU R & D on removal of codeposited layers and tritium, in:Fusion
Technology 1998, Proceeding of 20th
SOFT, Marseille, France, 1998, pp. 185188[6] W. Jacob, B. Landkammer, C. H. Wu, J.
Nucl. Mater. 266-269, 552(1999)[7] H. Kwast, H. Werle, C. H. Wu, Physica
Scripta Vol. T64, 41(1996)[8] C. H. Wu, J. P. Bonal, H. Kwast, F.
Moons, G. Pott, H. Werle, G. Vieider,Fusion Engineering and Design, 39-40,263(1998)[9] D. Boyer, Ph. Cetier, L. Dupin, Y. Belot,
C. H. Wu, Tritium sorption incarbon dust, in: Fusion Technology 1996,
Proceeding of 19th SOFT, Lisbon,
Portugal 1996, pp. 1751-1754[10] A. T. Peacock, P. Cetier, C. H. Wu et al, J.
Nucl. Mater. 266-269, 423(1999)

Acknowledgements
The author would like to thank A. T.
Peacock, W. Eckstein, J. Roth, H. Werle, A.A.
Haasz, G. Fussmann for useful discussion andcontribution in frame of EU fusion R & D
Program.