L.B. Begrambekov1, O.I. Buzhinsky2, A.M. Zakharov1

 

1National Research Nuclear University MEPhI, 31 Kashirskoe sh., Moscow 115409, R.F2Troitsk Institute for Innovations and Fusion Research, Troitsk, 142190, R.F.

Possibility of protecting renewable B4C coating application in ITER.

19th International Conference on Plasma Surface Interactionin controlled fusion devices

San Diego, California. May 24-28, 2010

OUTLINE

1. Introduction

2. Problems of tungsten divertor tiles

3. Problems of carbon (CFC, graphites) divertor tiles

4. B4C as a renewable protecting coating for the tiles.

5. Hypothetic C-W-B4C tile

Tungsten

Tungsten . BLISTERING(cones and pyramids)

Fig.1. Blisters (cones and pyramids) on the W surface irradiated by high-flux (1022m-2s), high-fluence (up to 1027m-2) and low-energy (38 eV) deuterium plasma at T=480-520 K [1] . a) Small blisters, b) cavities inside them, c) pyramid-like big blister, d) cone-like big blister,

The temperature of blistering is specific for cooper irradiated with moderate fluxes.(T=480-520 K)

[1] W.M.Snu, M Nakamichi, V.KH. Alimov, G.-N. Luo, K. Isobe, T. Yamanishi et al. J. of Nucl. Mater. 390-391(2009)1071 .

Tungsten : Cone growth under irradiation with moderate ion flux

Ar+ , Ei =400 eV, j= 1,2×1020m-2s, Φ= 5×1022m-2, T=1150 K

Tungsten : BLISTERING

Рис. Blisters on the W surface irradiated in deuterium plasma E =90 eV, Φ = 3,4 · 1025 м-2, T= 550К [2]

The temperature of blister appearance is specific for cooper irradiated with moderate fluxes. (T=480-520 K)

2. Ohno 2007

Tungsten : FLAKING

Fig.1. Flakes on the W surface irradiated by high-flux (1022m-2s), high-fluence (up to 1027m-2) and low-energy (38 eV) deuterium plasma at the temperature T= 618 K [1].

The temperature of flakingis specific for cooper irradiated with moderatefluxes (T=618 K)

[1] W.M.Snu, M Nakamichi, V.KH. Alimov, G.-N. Luo, K. Isobe, T. Yamanishi et al. J. of Nucl. Mater. 390-391(2009)1071.

Tungsten : MELTING OF FLAKS

Local melting of W surface irradiated by high-flux (1022m-2s), high-fluence (up to 1027m-2) and low-energy (38 eV) deuterium plasma. Surface temperature is 618 K. Estimated temperature of flake melting is 1300-1400 K

The temperature of flake melting Is estimated to be (T=1300-1400 K). It is the temperature range of cooper melting point

[1] W.M.Snu, M Nakamichi, V.KH. Alimov, G.-N. Luo, K. Isobe, T. Yamanishi et al. J. of Nucl. Mater. 390-391(2009)1071.

Tungsten : SUB THRESHOLD SPUTTERING

Irradiation parameters: Deuterium ions, Ei =5 eV, j=(1-2) 1021m-2s , Φ= (0.5-1)×1026m-2 , T ≥1250 K

Sputtering yield measured : Y=(1-1.5)×10-4 at/ion, at T=1450 K expected : Y=(1/3-1)×10-3 at/ion, at T=1450 K

[3] M.I.Guseva, V.M. Gureev, B.N. Kolbasov, C.N. Korshunov, Yu.V. Martynenko, V.B. Petrov, B.I.Khripunov. Letters in J. of Experimental and Theoretical Physics 77/7(2003)430

Increase of sputtering yield is foreseen, if grain dimensions are decreased (Y~r -1/2) or ion flux is increased

Activation of mechanism of ion stimulated desorption indicates on pasivation of tungsten surface

Main stages of the phenomenon are as follows: - accelerated diffusion of the interstitial W atoms towards the surface, - removal of atoms from the surface through mechanism of ion stimulated desorption

Tungsten tiles under high-flux and high-fluence irradiation

High concentration of implanted deuterium atoms occurring in W under high-flux and high-fluence irradiation results in weakening of interatomic forces.Surface layer of tungsten transformed in the material with new properties. Already observed and expected “new properties” are as follows:-high plasticity,-low melting point,-weaker electrical conductivity,-weaker thermal conductivity,-smaller surface binding energy,-lower sputtering threshold, etc.

Expected consequences of simultaneous neutron- and high-flux and high-fluence

plasma irradiation of tungsten tiles of ITER divertor

High concentration of implanted hydrogen atoms will hamper annihilation of neutron induced vacancies and will promote an increase of saturation concentration of neutron produced traps. It will lead to further increase of hydrogen concentration in the tiles and to activation of all relevant phenomena.

Expected consequences of simultaneous neutron- and high-flux and high-fluence

plasma irradiation of tungsten tiles of ITER divertor

High concentration of implanted hydrogen atoms will hamper annihilation of neutron induced vacancies and will promote an increase of saturation concentration of neutron produced traps. It will lead to further increase of hydrogen concentration in the tiles and to activation of all relevant phenomena.

As a result one can expect significant decrease of the life time of the tiles and increase of both tritium retention and dust production.

Expected consequences of simultaneous neutron- and high-flux and high-fluence

plasma irradiation of tungsten tiles of ITER divertor

High concentration of implanted hydrogen atoms will hamper annihilation of neutron induced vacancies and will promote an increase of saturation concentration of neutron produced traps. It will lead to further increase of hydrogen concentration in the tiles and to activation of all relevant phenomena.

As a result one can expect significant decrease of the life time of the tiles and increase of both tritium retention and dust production.

Adequate in situ renovation of sputtered parts of the tiles is practically impossible as well as W dust removal. Thus often breaks of ITER operation for tile replacement and dust removal will be necessary.

Possible solution of the problems of W tiles

The possible solution of the problems of W tiles – is exclusion of simultaneous irradiation of the tiles by both neutrons and plasma ions.

The neutron irradiation cannot be avoided.

Direct plasma influence on the tiles can be avoided through application of renewable protecting coating.

Carbon (CFC, graphites)

Carbon : SPUTTERING

“Chemical sputtering is numerously decreased under high flux irradiation” [4].“The rate of CFC tile erosion and consequence tritium accumulation in the redeposited carbon films and dust formation will be at least one order of magnitude smaller than it was assumed earlier” [5]

•J. Roth, R. Preuss, W. Bohmeyer, S. Brezinsek, et al. Nucl. Fusion 44 (2004) L21.

[4] J. Roth, R. Preuss, W. Bohmeyer, S. Brezinsek, et al. Nucl. Fusion 44 (2004) L21.[5] J.Roth, E.Tsitrone, A.Loarte, et al. J. of Nucl. Mater. 390-391(2009)1.

Carbon : Tritium retention in the bulk of CFC tiles

T/C ratio in the bulk of the tiles is expected to be ≈ (3-5)×10-7.

The total tritium content in the bulk of CFC tiles of ITER divertor

(50 m2) could be equal to (2-3)×1022 T.

Carbon : Hydrogen retention in redeposited carbon layer

Conditions of redeposition (Н/С) ratio in the deposited films

1. Redeposition in tokamaksJT-60Т10

JET

≤ 0.040.06 – 0.30.1 – 0.4

2. Laboratory experiments● Hydrogen atmosphere

● Assisting plasma irradiation, RT Еi ≤ 200 eV Еi ≥ 400 eV

Deposition: temperature, substrate Т○С I Stainless steel Graphite 100 I 0.20 I 0.10 220 I 0.18 I 0.09 430 I 0.05 I 0.06 630 I 0.03 I 0.02

100 I ≈ 0.2 100 I ≈ 0.4

3. H/C ratio earlier used for estimation of ITER T-limit

H/C ratio accepted now for estimation of ITER T-limit

0.4

0.2

Carbon tiles under neutron and plasma irradiation

Critical issue Today understanding

Chemical sputtering Erosion rate is small

Tritium accumulation in depositedcarbon films

Deposition rate is not high. Deposited films could be removed

Production of tritiated dust Could be avoided or reducted

Tritium accumulation in the tiles Retention in bulk of C-tiles is small.Deposition on the tiles could be removed

Replacement of sputtered parts of the tiles

Adequate replacement of sputtered parts of the tiles is impossibly

Tritium retention in the tiles under neutron irradiation

Tritium retention enhances

Tritium removal from the tiles Tritium removal under acceptable temperatures is impossible

Dimensional changes underneutron irradiation

Dimensional changes is not avoidable, if T≤ 800-900 K

Possible solution of the problems of carbon tiles

The promising solution of the problems of carbon tiles is

1) exclusion of direct plasma influence on the tiles through application of renewable protecting coating,

2) keeping of the tiles (or carbon parts of the tiles) at the temperature T≥800-900 K to provide annealing of neutron induced defects and to prevent dimensional changes

Boron carbide coating as a renewable protecting coating

for ITER divertor- and first wall tiles

Parameters of B4C. SPUTTERING

Fig.3. Temperature dependence of the erosion yield due to 1 keV D+ sputtering and evaporation for various

materials [6].

[6] J.Roth, J.Nucl.Mater. 176&177 (1990) 132-141

Parameters of B4C. HYDROGEN RETENTION

Fig.4. Temperature dependence of hydrogen retention for different materials (H+, Ei=100 eV, j=5.61019 m-2) [7].

. [7] L. Begrambekov, O. Buzhinsky, A. Gordeev, E. Miljaeva, P. Leikin, P. Shigin. Physica scripta, N108 (2004), p.72-75.

Parameters of B4C. Behavior under high power irradiation

Temperature of B4C coating on fine grain graphite MPG-8 and high thermal conductive graphite RGT under high power electron irradiation[1].

Heat flux, MW/m2

Time,sec

Number of cycles

Тmax, 0C

MPG-8 RGT

2,02,3

55

2020

< 300-

-< 300

5,05,8

2,52,5

1010

430-

-< 300

5,05,8

1010

1010

940-

-790

10,011,013,0

255

255

> 1400-

-880940

[8] Magnetic fusion energy program, annual report SNL(1989)18.( reconstruction of tables 1 and 2)

Parameters of B4C. The product of B4C sputtering

D/C ratio in sputtered boron carbide coating

D/C ratio in redeposited (H/C/B) layer

Comments

4.12 0.22-0.45 Redeposited layer could be removed in cleaning discharge

Laboratory experiment

Parameters of B4C. In situ coating deposition

Deposition conditions Plasma providing total dissociation of the molecules of initial substance

Devices used for B4C deposition

Tokamak T-11M, PISCES-B (Te~40 eV, ne~2·1017m–

3, electron flux~2·1017m –2с–1) [9].

Initial substance Non-toxic, non-explosive, and non-hazardous carborane (C2B10H12)

Deposition rate ≈ 30 nm/s (1µm/min) in PISCES-B discharge

[9] Buzhinskij O.I., Otroschenko V.G., Whyte D.G. et al. J. Nucl. Mater., 313—316 (2003) 214.

Advantages of B4C coating and further investigations

● B4C coating will provide low rate erosion of divertor tile surfaces.

● B4C coating will prevent erosion of the tiles as well as penetration of tritium into and trapping in the bulk of the tiles.

● B4C coating can be deposited and renewed during regular ITER discharges

● B4C coating can withstand high thermal fluxes (13.0 M∙W∙m-2)

● Erosion of B4C coating will lead to deposition of easily outgased and easily removed H/C/B films.

● Estimation shows that tritium concentration in the redeposited films approached T-limit and removal of the films will be needed after ≥ 2500-3000 ITER discharges (400s, Q=10).

● Expanded investigation of deposition and behavior of B4C coating in tokamak conditions is needed

Hypothetic C-W-B4C tile

Scheme of B4C-W-C tile

1. Water cooled pedestal, 2. Cooling water, 3. Tungsten plate 4. Tungsten sticks, 5 . RGT graphite or CFC, 6 . B4C coating

Temperatures of outer and inter boundaries of the tile under14 MW/ m2 irradiation (λW=120 W/mK, λB4C=20 W/mK, λРГТ=200 W/mK)

Thickness of the parts TemperatureOverall thickness of the tile 10 mm Temperature of plasma facing surface 1400⁰C Thickness of B4C coating 50 µm Temperature of B4C – С boundary 1360⁰C

Thickness of RGT plate 4 mm Temperature of C – W 900⁰C Thickness of tungsten sticks 5 mm Temperature of W – pedestal 200⁰C Thickness of tungsten plate 1.5 mm

3

1

2

4

5

6

The coating protects tile materials from erosion and particle implantation.

Tungsten parts are not subjected to influence of accidental off-normal events.

An amount of sticks is selected to keep temperature of graphite plate above 800-900 K providing annealing of neutron induced damages.

Conclusion

Number of phenomena observed last years allows saying that high-flux and high-fluence plasma irradiation generates a high quasi-equilibrium concentration of implanted hydrogen particles in W. It results in weakening of interatomic bonds and modification of W properties in a wide temperature range. Activation of mentioned processes under neutron irradiation will remarkably influence the critical plasma wall interaction issues for ITER. High-flux and high-fluence plasma irradiation does not influence hardly carbon material properties Nevertheless they cannot be used as a plasma facing materials at the activated stage of ITER. The property of boron carbide coating as a plasma facing material is considered. Conclusion is made that boron carbide layers can be used as a protecting renewable coating for divertor tiles at the activated stage of ITER The scheme of hypothetic C-W-B4C tile is discussed. Assumption is made

that application of the tiles of such types allosw increasing the duration of ITER operation without replacement of the divertor tiles.

Thank you for your attention!

Estimation of carbon tile behavior in ITER

“… in some cases the data at the extreme upper limit were used for estimation of carbon tile behavior in the ITER divertor.”

[ ] J.Roth, E.Tsitrone, A.Loarte, et al. J. of Nucl. Mater. 390-391(2009)1.

IntroductionHigh-flux and high-fluence irradiation leads to number of phenomena which

have never been observed or been observed in different temperature ranges.

Among them there are formation of blisters and bubbles under low energy ion irradiation , flakes and their low temperature melting, blisters and blister-like features (which are cones, pyramids and hills), fuzzing, sub-threshold sputtering of tungsten and decrease of chemical sputtering of graphites.

These phenomena being considered in total help to reveal general regularities of the specific processes in the materials subjected to irradiation with high-fluxes and high-fluence irradiation.

The report briefly analyses behavior of both tungsten and carbon (CFC) under high-flux and high-fluence irradiation. The possible consequences of their appearance in the ITER simultaneously with n-irradiation are discussed.

Boron carbon (B4C) coating is considered to be the way to prevent their negative influence on critical plasma wall interaction issues for ITER. In conclusion the scheme of hypothetic C-W-B4C tile is discussed