Progress of the European R&D on plasma–wall interactions, neutron effects and tritium removal in ITER plasma facing materials

Fusion Engineering and Design 56–57 (2001) 179–187

Progress of the European R&D on plasma–wallinteractions, neutron effects and tritium removal in ITER

plasma facing materials

C.H. Wu a,*, C. Alessandrini b, J.P. Bonal c, J.W. Davis d, A.A. Haasz d,W. Jacob e, A. Kallenbach e, J. Keinonen f, P. Kornejew g, R. Moormann h,

V. Philipps i, J. Roth e, F. Scaffidi-Argentina i, H. Wurz i

a EFDA, Max-Planck Institut fur Plasmaphysik, Boltzmannstraße 2, 85748 Garching, Germanyb ENEA Sulla Fusione, C.R.E. Frascati P. O. Box 65-00044, Frascati, Rome, Italy

c CEA, Saclay, 91191 Gif-sur-y�ette, cedex, Franced Uni�ersity of Toronto, UTIAS, 4925 Dufferin Street, Toronto, Ont., Canada

e Max-Planck Institut fur Plasmaphysik, 85748 Garching, Germanyf Accelerator Laboratory, P. O. Box 43, Uni�ersity of Helsinki, FIN-00014 Helsinki, Finland

g Max-Planck Institut fur Plasmaphysik, D-10117 Berlin, Germanyh Forschungszentrum Julich, D-52425 Julich, Germany

i Forschungszentrum Karlsruhe, 76021 Karlsruhe, Germany

Abstract

In a next step D/T fusion device like ITER, an intense neutron flux will be produced as a consequence of thenuclear fusion reactions. The effects of the neutron induced damage in the microstructure of the plasma-facingmaterial (PFM) may significantly change the thermal properties and the mechanical properties as well as thebehaviour of the swelling and the tritium retention in such materials. In addition, a peak heat flux as high as 20 MWm−2 and a plasma flux of 1018–1020 cm−2 s−1 are expected in the divertor zone during the normal operation of thereactor. The divertor materials have to withstand the neutron damage, the high heat fluxes and the high erosioncaused by the interaction with the high flux plasma. The sputtered particles are co-deposited with plasma, which maycontribute significantly to the total tritium inventory in the PFM. Furthermore, the interaction of steam with thesputtered particles (with usually high specific surfaces) could produce large amounts of hydrogen. All of the abovetopics represent critical issues for plasma performance, safety and economy, as they could limit the use of some PFMmaterials in next generation fusion devices. Therefore, substantial R&D effort is needed to elucidate the effects of theneutron induced damage on microstructure, erosion/deposition, tritium retention and dust formation, as well as onhydrogen production. In the framework of the European Fusion R&D program, an extensive effort on neutron effectsof the material properties: namely, thermal conductivity, mechanical properties, dimensional stability, tritiumtrapping, erosion/deposition, co-deposition, dust formation/removal, chemical reactivity with steam and oxygen,outgassing, baking and tritium removal from PFM have been undertaken during the past several years. In this paper,

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C.H. Wu et al. / Fusion Engineering and Design 56–57 (2001) 179–187180

the recent progress achieved within the European Fusion R&D program and contributions to the development ofITER PFMs are presented and critically discussed. © 2001 Elsevier Science B.V. All rights reserved.

Keywords: European R&D; Plasma–wall interactions; ITER plasma facing materials

1. Introduction

Armour materials currently proposed for ITERare beryllium, carbon-fibre-composites and tung-sten. Among the low-Z materials, carbon andberyllium are seriously considered as plasma fac-ing materials (PFMs) for the next step FusionExperimental Reactor. This is partly due to theextensive experience with these materials for firstwall and divertor plate protection in present toka-maks. In addition, excellent plasma performancehas been demonstrated with C and Be. Carbonbased materials have been chosen for protectionof high heat flux components, in particular, foroff-normal operation (e.g. disruption, plasma ex-cursion), whilst beryllium has been proposed as aPFC material for gettering the oxygen impurity inplasma. Tungsten (W) has a high physical sputter-ing threshold, which leads to low erosion: hence, alonger component lifetime is expected with W.Erosion of the divertor and first-wall plasma-fac-ing components, and tritium uptake in the rede-posited films on the armour material surfacessurrounding the plasma, represent crucial physicalissues that impact on the design of the next stepdevice. In addition, as next generation D/Tplasma devices, e.g. ITER, will produce intenseneutron fluxes, substantial R&D is needed tostudy the effects of neutron-induced damage onthe microstructure and critical properties of thesematerials (e.g. thermal conductivity, swelling, andtritium trapping), because they could limit the useof these materials in next generation fusiondevices. Neutron induced changes in thermal con-ductivity, dimensional stability, mechanical prop-erties as well as the behaviour of tritium–wallinteractions are crucial problems that need to bebetter understood. The issues of erosion, co-depo-sition and neutron effects are being explored byboth dedicated experiments in laboratory simula-tions and modelling. This paper presents the re-sults of European Fusion R&D on plasma–wall

interactions, neutron effects as well as on theremoval of tritium and co-deposited layers inplasma facing materials planned for ITER.

2. Plasma–wall interactions

2.1. Erosion and redeposition at di�ertorconditions

2.1.1. Flux dependence of chemical erosionCarbon chemical erosion yields have been mea-

sured by means of spectroscopic particle flux mea-surements and with Langmuir probes in the outerDivertor II of the ASDEX Upgrade tokamak [1][2]. The molecular hydrocarbon flux is derivedfrom CH band spectroscopy, employing a re-cal-culation of the original hydrocarbon (CH4) releasefrom the measured flux of the CH breakup prod-ucts using a fixed loss factor for the fast redeposi-tion of intermediate products taken from simpleanalytical modelling. The deuterium ion fluxes aretaken from Langmuir probe measurements at theintersection point of the viewing line with thedivertor target, which were cross-checked withhydrogen influx measurements via H�

spectroscopy.The chemical erosion yields obtained from the

recalculated hydrocarbon wall flux and theplasma ion flux exhibit an isotope effect andpronounced flux density dependence (Fig. 1):

Ychem=1.4·MH [amu]·�H−0.75 for flux

�1022 m−2 s−1.

Where M is the mass of isotope and � is the fluxdensity.

2.1.2. Flux and target temperature dependence ofchemical erosion

The plasma generator PSI was used for anextensive series of experiments in steady stateplasmas to investigate the chemical erosion in thehigh flux range during the last few years. The

C.H. Wu et al. / Fusion Engineering and Design 56–57 (2001) 179–187 181

Fig. 1. Chemical erosion yields as a function flux density(ASDEX-Upgrade).

Fig. 2. Chemical erosion yields as a function of temperature, atflux density 1.8×1023 m−2 s−1 (Plasmagenerator).

details of apparatus and experimental procedureswere described elsewhere [3–5]. The ion flux den-sity was varied between 1.4×1022 and 1.8×1023

m−2 s−1, the electron temperature Te was kept atabout 5–10 eV, in order to realise a particleimpact energy onto the samples of less than 30eV, i.e. below or near the threshold for physicalsputtering. Higher density values are usually cor-related with the higher temperatures. The mea-surements of chemical erosion were carried outwith optical spectroscopy (CD-band around 430nm), mass spectrometry using a differentiallypumped quadrupole mass spectrometer and bythe weight loss measurements. For comparison,carbon fibre composite (CFC)-samples of CON-CEPT II (DUNLOP) and silicon-doped NS 31(SEP) were exposed to hydrogen and deuteriumplasmas after appropriate cleaning and outgassingprocedures. The target temperatures were between250 and 900 °C. The in situ methods—opticalspectroscopy and mass spectrometer—were cali-brated for each ion flux density with the knownfluxes of CH4 (CD4), the main product of chemi-cal erosion. Therefore, for evaluation, no assump-tions concerning photon efficiencies and thefraction of CH (CD) species reaching the massspectrometer ion source are necessary. The mea-sured chemical erosion yields as a function of ionflux density were reported elsewhere [4]. It is seenthat the chemical erosion yields decreased withincreasing ion flux density. At the highest ion flux

density (1.8×1023 m−2 s−1), the yield of chemi-cal erosion is as low as around 0.2–0.5%, depend-ing on target temperature (Fig. 2). The resultsalso show, that the chemical erosion yield ofSi-doped CFC is lower than that of the undopedCFC.

2.1.3. Molecular dynamics simulation on fluxdependence of chemical erosion

The results of non-cumulative simulationsdemonstrated a dramatic difference in the carbonchemical erosion yields between bombardment ofunsaturated and supersaturated a-C:H layers. Thecarbon yield for the unsaturated surface is �0.01,while for the supersaturated surface is as low as�0.001. This reduction by an order of magnitudeis due to the high hydrogen content leading to adecreased carbon collision cross section, namely ashielding of carbon by the hydrogen atoms. Tri-tium simulation showed that the reflection of theincident ions dominated T sputtering with boththe unsaturated and supersaturated surfaces. Thetritium yield per incident ion with an unsaturatedsurface is 0.54, of which 65% were reflected inci-dent ions and 26% sputtered T2 dimers. For asupersaturated surface the tritium yield was muchhigher, very close to 1.0, and the fraction ofreflected ions had decreased to 47%, while the


sputtering of T2 molecules had increased to 45%of all the etched tritium [6,7].

2.2. Modelling, experimental �alidation ofoff-normal e�ents

Extensive validation of FOREV-2 using theresults from simulation experiments performed atthe MK-200 UG facility has shown that turbulentprocesses in the plasma shields of horizontal andvertical targets are absent [8]. The plasma shielddynamics and its stability can be described ade-quately by the classical Spitzer diffusion coeffi-cient. Moreover, the validation exercise hasdemonstrated that the properties of the plasmashields produced in the simulation experimentsare rather comparable to those produced in toka-mak disruptions. The only difference is the largeratomic density in the tokamak case. Therefore theconclusions on quality of calculated parametersare also valid for the tokamak case. The validatedFOREV-2 was applied to calculate erosion ofITER-FEAT vertical graphite and tungstentargets and to calculate transport of impuritiesduring off-normal events and ELMs from thedivertor towards the central plasma. Flaking fromredeposited layers and macroscopic erosion dur-ing disruptions produce complex layers of consid-erable surface roughness and drastically changedthermophysical properties. The hot spots of suchlayers are responsible for enhanced impurity pro-duction. First numerical estimations on maximumtolerable ELM energy show that the tolerableenergy is considerably lower for such layers thanfor the virgin target and might be as low as 0.5MJ m−2 [9]. Melt layer erosion of metals isdominated by melt flow. The erosion value is upto a factor of 2 larger than the melt thickness.Erosion of tungsten FWs can be up to 2 mm, forBe up to 3.5 mm for RAE impact. Due to thelarge damage for RAEs it is mandatory to miti-gate those events or to limit the tolerable RAEenergy density to values below 20 MJ m−2. Meltlayer erosion is always accompanied by splashing.Up to 20% of the eroded mass is splashed awayby droplets. The multifluid code FOREV-2 allowsa detailed 2-D modelling of plasma–wall interac-tion under off-normal conditions and allows quite

a realistic damage analysis of erosion byevaporation.

2.3. The influence of silicon on the characteristicsof CFC

To improve the properties of carbon materials,one strives for reduced tritium inventory, reducedchemical erosion and increased resistance to wa-ter/oxygen at elevated temperatures. In addition,high heat loading due to off-normal events (e.g.plasma disruption, slow transients, and ELMs,which can occur during a transition from de-tached to attached divertor operation) requireshigh thermal conductivity for the carbon protec-tion material. Therefore, CFCs with high thermalconductivity (300 W m−1 K−1, at 20 °C, 145 Wm−1 K−1, at 800 °C) are favourable. In theframework of the European Fusion Technologyprogram, a great effort has been made to developCFCs to meet all of these requirements. In partic-ular, Si doped CFCs-SEP NS31 were developed,and their properties have been investigated. NS31is a 3D CFC constituted of P55 ex-pitch fibres(80% vol.) in one direction and ex-PAN fibres(20% vol) in the perpendicular direction; this CFCundergoes a subsequent needling which gives thefibres orientation in the third direction (z direc-tion). The high thermal conductivity direction isthat of the P55 ex-pitch fibres. NS31 is densifiedby chemical infiltration of pyrocarbon and heattreated at a temperature �2500 °C. At last,liquid silicon is injected under pressure leadingpartly to the formation of silicon carbide (SiC).NS31 contains about 8–10 at% of silicon and itsporosity is about 3–5% [10].

2.3.1. Chemical erosionThe chemical erosion yield of the NS31 mate-

rial was lower by a factor of about 2–3 in com-parison with undoped graphite [4,11].

2.3.2. Oxidation beha�iour of CFCs and dopedCFCs

Oxidation resistance is an important criterion inthe selection of PFCs. PFC-oxidation may havesevere consequences in case of loss of vacuum,and loss of coolant into vacuum accidents (water/


steam ingress into the vacuum vessel). Details ofexperimental investigations of doped carbon (Ti,Si) and doped CFCs exposed to steam and air aredescribed elsewhere. Comparing steam and air, itwas found, that oxidation rates in air at 700 °Care roughly about the same as the rates in steamat about 1000 °C. In general, undoped CFCs havea higher oxidation rate. Highly SiC doped INOXA14 (40% vol. SiC) revealed an oxidation resis-tance in steam about 2–3 orders of magnitudebetter than undoped materials; reduction of Sicontent to 8–10 at.% decreases the oxidationresistance in comparison with INOX A14(40%vol. SiC), which, however, still remains betterthan for undoped carbons by a factor of about2–4 [12]. Doping with Ti, however, does notimprove oxidation resistances.

2.3.3. Outgassing beha�iour of CFCsA systematic investigation of outgassing be-

haviour of doped CFCs and undoped CFCs wascarried out in terms of temperature dependenceand temperature pre-treatment. The experimentaland evaluation procedures have been described indetail elsewhere. For comparison, the characteris-tics of five materials, SEPcarb N112 (3D), DunlopV(2D), SiC 2.5%, SiC 8% doped CFCs (2D), 3DNovotex (No3) and NS31 (Si-doped 3D) havebeen investigated. In order to obtain appropriatecomparison, all pre-treated samples (1000 °C, 20h) were exposed to air for 2 weeks before theoutgassing test. In general the total amount ofoutgassing of Si-doped CFC is 1–2 orders ofmagnitude lower than that of other carbon basedmaterials tested. The results also indicate that therelease temperature of H2O of Si-doped CFC isabout 100 °C lower than other carbon basedmaterials [13].

3. Removal of co-deposited layers and tritium

3.1. Remo�al by O2 and ozone

A series of experiments has been performed onthe thermo-oxidation of tokamak codeposits fromvarious fusion reactors [14–16]. In Ref. [14], acomparison was made between the oxidation of a

thick (�5 �m) TFTR codeposit and a laboratoryfilm. Erosion yields of more than 1000× largerwere observed for the tokamak film, which wascompletely removed by �1/2 h at 623 K in 2.1kPa O2 (Y�6×10−6 C/O2). The oxidation yieldfor a codeposit obtained from JET could not bedetermined without a measure of the film density,however, erosion rates of �3.3 �m h−1 wereobserved [15]. This is about a factor of 3 smallerthan that seen for the thick TFTR specimen[14,15]. The erosion of the DIII-D codeposits [15]was similar to the JET specimen, but the filmdensity could be estimated in this case, and yieldswere determined to be �2×10−6 C/O2 at 623 K.The erosion rate was found to increase with in-creasing pressure (2.1–21 kPa), but less than lin-early, so that yields decreased with increasingpressure. More recent specimens from DIII-D,which have been exposed to many boronisationprocedures, behaved quite differently. Thermo-ox-idation of midplane codeposits which were about2 �m thick, but containing mostly boron, revealedthat only about 1/4 of the D could be removed byO2 at 623 K [16]. Presumably, the boron contentof the film was forming a stable oxide, halting thefilm removal process. On specimens from both theupper and lower divertor, the D concentration islower, and somewhat more readily removed; how-ever, it is not clear that the deuterium is in asurface film, as opposed to an implanted layer.

TEXTOR flakes (thickness 20 �m, D/C=0.04,metal and boron content: 15 weight-%) howevershow initial removal rates of only 0.07 resp. 0.25�m h−1 at 573 resp. 623 K in oxygen (21 kPa);the rate slows down strongly with burn-off. Acomplete C-removal is possible at 773 K. A labo-ratory produced very homogeneous a-C:D layer(DIARC, 3.5 �m of thickness, D/C=0.1) is oxi-dised in oxygen (0.3–20 kPa) at temperatures of523–723 K in order to generate D-releases. Mainresults show that a minimum temperature valuesuitable to produce D release (mainly in form ofHDO) is 623 K, oxygen partial pressure 0.55 kPa[17]. SIMS analysis reveals a simultaneous releaseof C and D. a-C:D-flakes from TEXTOR wereinvestigated in ozone/oxygen mixture (2.5%ozone); at 180 °C the oxidation started withweight loss rates of about 10% h−1, but the rate


slows down, when one third of the a-C:D flakes isreacted [17,18].

3.2. Remo�al by ECR discharge

The erosion of soft and hard C:H layers byhydrogen ECR plasmas was investigated in greatdetail in the past [19,20]. Later on, the erosionrates of soft and hard C:H layers using differentworking gases (O2, D2, H2, O2/H2 mixtures, andH2O) were determined [20]. In all investigatedcases, the erosion in pure oxygen plasmas yieldedthe highest rates. The erosion rates increase sig-nificantly by up to a factor of 10 if a bias voltageof up to 250 V is applied [19]. If no bias voltageis applied, the erosion rates can be increased byup to a factor of 4 if the surface temperature israised from 300 to 650 K [20]. The role of ionsand neutrals in this erosion process was alsoinvestigated. It was shown that the erosion rate isdirectly proportional to the energy deposited intothe layers by the impinging ion flux [20].

In all cases, oxygen always shows the highestrate. At T=300 K the relative rates behave ap-proximately as O2:D2:H2=10:2:1.The highestrates achieved in oxygen discharges at 300 K withthe substrate at floating potential are 3.6 �m h−1

(=1 nm s−1) for polymer-like, soft C:H films(H/C about 1) and 1.7 �m h−1 (=0.48 nm s−1)for hard a-C:H films (H/C about 0.5). The erosionrates increase with increasing substrate tempera-ture (roughly by a factor of 3–4, if T is raisedfrom 300 to 650 K) and with increasing ionenergy (by a factor of about 6–8, for 0–200 eV).

4. Neutron irradiation effects

4.1. Carbon

Investigations on neutron irradiation effects onthe thermal–mechanical properties of severalCFCs have been carried out. Three main irradia-tion experiments were performed, namely, CE-RAM D217-15, CERAM D217/18/19 andMACIF. The irradiation temperatures were:CERM D217-15, 1500 °C; CERAM D217/18/19,600, 800, 1000 °C; and MACIF, 400 °C. The

dpa.g values were calculated by the HFR-TEDDIprogram. The CERAM irradiations have beenperformed in the High Flux Reactor (HFR),Petten, whilst the MACIF irradiation has beencarried out in the OSIRIS reactor at Saclay(France). The damage dose of the Petten HFRand MACIF irradiations ranged from 0.8 to 5and 0.35 to 0.85 dpa.g, respectively. The charac-terisation of the latest EU irradiation programmewas performed in 1998. The irradiation conditionswere 0.3/0.35 dpa.g at 335 and 775 °C,respectively.

4.1.1. Neutron effects on thermal conducti�ityNeutron irradiation induces degradation of

thermal conductivity. For example, after irradia-tion at a temperature of 420 °C and a damagedose of 0.83 dpa.g, a degradation of thermalconductivities has been observed for all investi-gated CFCs. The normalised thermal conductivi-ties at 400 °C, (Ki/K0)400°C, were between 0.28 and0.38. Neutron irradiation induces carbon displace-ment from their initial positions in the lattice tointerstitial positions between two basal planes.Thus, large dislocation loops or defect clusters arecreated especially at low irradiation temperatures.The low irradiation temperature (400 °C) preventsannealing of these defects, and the propagation ofthe phonons along the basal plane is sloweddown, inducing a decrease in the thermal conduc-tivity. The average normalised thermal conductiv-ities of all investigated CFCs as a function ofirradiation temperature show that the neutroninduced degradation of thermal conductivity de-creases with increasing irradiation temperature(Table 1). There is almost no degradation ofthermal conductivity, if the irradiation tempera-ture is higher than 1000 °C for neutron damage of1 dpa.g. It is obvious that there is a beneficialeffect of high irradiation temperature (�1000°C), allowing annealing to occur during irradia-tion, so that the carbon atom displaced from theirlattice sites can return to their initial position. Theresults also show that the neutron induced degra-dation of thermal conductivity is lower for thoseCFCs which have lower initial thermal conductiv-ities [21,22].


4.1.2. Dimensional changesThe dimensional changes (dL/Lo) of 2D and

3D CFCs irradiated in the temperature ranges385–420 °C, to 0.41–0.83 dpa.g, are very small inboth parallel (�) and perpendicular (�) directions.Generally, dimensional changes of ex-PAN fibresCFCs are negative in both directions. In contrast,the dimensional changes of ex-pitch fibres CFCsare always positive in both directions and are veryclose to zero. After irradiation at 600 °C dpa.g−1,the dimensional changes are almost the same as at400 °C [23].

4.1.3. Neutron irradiation effects on tritiumretention in carbon

Before tritium loading, the unirradiated speci-mens were conditioned for 1 h at 850 °C undervacuum to remove air and moisture. For theirradiated specimens the conditioning temperaturewas adapted to the irradiation temperature. Tri-tium loading was done in an H2+2 ppm T2 gasmixture. After loading, the surface adsorbed tri-tium was removed by purging at room tempera-ture. The retained tritium was determined byheating the specimens at a rate of 5 °C min−1 toa temperature of 1050 °C and purging with He+0.1% H2. The released tritium is measured with anionisation chamber and finally trapped in bub-

blers [24]. It was observed that about 10% of theretained tritium was not released by heating to1050 °C. To determine the total retained tritiumthe samples were normally oxidised afterwards inoxygen at 850 °C. The tritium trapped in thebubblers was determined by liquid scintillationcounting. The most systematic investigation ofneutron effects has been carried out for graphiteS1611. The results show that the tritium retentionincreases with increasing neutron damage and itseems to achieve a saturation value about �0.1dpa. The fact that equilibrium is obtained forboth unirradiated and irradiated samples within10 h at 850 °C loading, indicates fast tritiumtransport and fast adsorption kinetics. Moreover,because essentially all trap sites should be occu-pied, the retained tritium should correspond tothe total number of trap sites. From the dataobtained, it is reasonable to estimate that thesaturation value (�0.1 dpa) of trap sites of neu-tron irradiated samples should be about 1000appm [24,25]. Recent results showed that Si-dop-ing increased the tritium release rate.

4.2. Beryllium

4.2.1. Neutron irradiation effects on the tritiumretention in beryllium

The samples studied were manufactured byBrush–Wellman (material type S200 HIP) andirradiated at Mol BR2. The loading was per-formed in an H2/T2 gas mixture (H2/T2=5000�30%). For comparison, the irradiated andunirradiated samples were loaded together in onerun. Tritium was released by purging with He+0.1%H2 and heating the samples slowly with 5 °Cmin−1 or 26 °C min−1 to a temperature of �850°C. From the tritium activity measured with anionisation chamber, the release rate and totalamount of released tritium were determined. Theresults show that neutron irradiation led to in-creasing tritium retention in beryllium by factorsof 15–43 at the experimental conditions. Releaseof neutron generated tritium by thermal annealingrequires high temperatures (850–1050 °C). Re-lease kinetics depend, among other factors, on thedensity, microstructure, the content of BeO andirradiation conditions. For irradiation with a

Table 1The normalised thermal conductivity (Ki/K0)T irr

as a functionof irradiation temperature

(Ki/K0)T irr

CFCs irradiated at 1 dpa.gIrradiation

Low K0 CFCbHigh K0 CFCaTemperature (°C)

0.30/0.35 0.30/0.354000.60/0.650.55/0.606000.85/0.90800 0.75/0.800.95/11000 0.70/0.85

1500 11

a K0 at 25 °C300 W m−1 °C−1, K0 at 800 °C120 W m−1

°C−1.b K0 at 25 °C100 W m−1 °C−1, K0 at 800 °C50 W m−1

°C−1.Ki is the thermal conductivity after irradiation, and K0 is thethermal conductivity before irradiation.


Fig. 3. Release of tritium and helium from neutron irradiatedBe at 1100 °C.

occurred. SEM photographs showed clear evi-dence of blister formation at the surface. The ionenergy of 500 eV/D+ is not high enough to pro-duce damage in the tungsten lattice, except possi-bly at substitutional impurity sites. It was thussuspected that the blisters resulted from the pre-cipitation of deuterium molecules at grainboundaries, or other natural defects in thematerial.

6. Conclusion

1. The results of ASDEX-Upgrade and Plasma-generator show that the chemical erosionyields decrease with increasing flux density.

2. The effects of Si-doping:� The chemical erosion yield is decreased in

comparison with undoped CFCs.� The release temperature of H2 and H2O of

Si-doped CFC (NS31) is about 100 °Clower than that of undoped carbon basedmaterials.

� Doping of Si/SiC (NS31) increases the oxi-dation resistance.

3. Removal of co-deposited layers and tritium:� The results of systematic investigations

showed the co-deposited layers can be effec-tively removed via interaction with oxygenor ozone.

4. Neutron effects:Carbon:� Detailed investigation of neutron induced

dimensional changes, degradation of ther-mal conductivity and tritium trap sites hasbeen performed for neutron irradiationtemperatures between 400 and 1500 °C andneutron damage between 10−3 and 12dpa.g.

� At an irradiation condition 1000 °Cdpa.g−1 3D CFCs showed better dimen-sional stability than 2D CFCs.

� The normalised thermal conductivity at theirradiation temperature, Ki/K0, increaseswith increasing irradiation temperature be-tween 400 and 1000 °C.

lower fluence (�1021 cm−2), a temperature ofabove 1000 °C is needed to release all the tritiumwithin several hours. Release from highly irradi-ated beryllium (�3×1022 cm−2) is faster, andalmost all tritium is released at 850 °C withinseveral hours. The results show that the neutroninduced saturation value of trap sites is around10,000 appm [26,27]. Moreover, a concurrent re-lease of tritium and helium from neutron irradi-ated Be was observed in a recent study (Fig. 3),which may lead to the conclusion that tritium andhelium reside in common bubbles [28].

5. Blistering of high-Z material tungsten

The formation of blisters on high-heat-flux sur-faces in fusion reactors would result in rapidoverheating and vaporisation of the outer layer,due to poor thermal contact with the bulk mate-rial. This could lead to contamination of the coreplasma, and it is important to employ materialsunder conditions for which this is unlikely. Evi-dence of blister formation on tungsten has beenobserved in ion-beam experiments. During experi-ments on D+ implantation in polycrystallinetungsten foils [29], it was observed that previousimplantations led to increased retention in foilsimplanted above 300 K. This led to a detailedstudy of the influence of ion damage on deuteriumtrapping in polycrystalline tungsten [30]. On spec-imens implanted at 500 K to fluences �1024 Dm−2, visible modification of the implanted surface


� Neutron induced tritium trap sites increasewith increasing neutron damage. However,it seems to reach a saturation level of 1000appm at a neutron damage of 0.1 dpa.g.Beryllium:

� To release the neutron generated tritium, atemperature �850 °C is required.

� It seems that the neutron induced satura-tion value of trap sites is around 10000appm.

5. Evidence of blistering on tungsten has beenobserved in ion-beam experiments, which mayalso effect H/D retention.

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Documents

Progress of the European R&D on plasma–wall interactions, neutron effects and tritium removal in ITER plasma facing materials