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L.B. Begrambekov 1, O.I. Buzhinsky 2, A.M. Zakharov 1 1 National Research Nuclear University MEPhI, 31 Kashirskoe sh., Moscow 115409, R.F 2 Troitsk Institute.

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Presentation on theme: "L.B. Begrambekov 1, O.I. Buzhinsky 2, A.M. Zakharov 1 1 National Research Nuclear University MEPhI, 31 Kashirskoe sh., Moscow 115409, R.F 2 Troitsk Institute."— Presentation transcript:

1 L.B. Begrambekov 1, O.I. Buzhinsky 2, A.M. Zakharov 1 1 National Research Nuclear University MEPhI, 31 Kashirskoe sh., Moscow 115409, R.F 2 Troitsk Institute for Innovations and Fusion Research, Troitsk, 142190, R.F. Possibility of protecting renewable B 4 C coating application in ITER. 19 th International Conference on Plasma Surface Interaction in controlled fusion devices San Diego, California. May 24-28, 2010

2 OUTLINE 1.Introduction 2.Problems of tungsten divertor tiles 3. Problems of carbon (CFC, graphites) divertor tiles 4.B 4 C as a renewable protecting coating for the tiles. 5.Hypothetic C-W-B 4 C tile

3 Tungsten

4 Tungsten. BLISTERING (cones and pyramids) Fig.1. Blisters (cones and pyramids) on the W surface irradiated by high-flux (10 22 m -2 s), high-fluence (up to 10 27 m -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.

5 Tungsten : Cone growth under irradiation with moderate ion flux Ar +, E i =400 eV, j= 1,2×10 20 m -2 s, Φ= 5×10 22 m -2, T=1150 K

6 Tungsten : BLISTERING Рис. Blisters on the W surface irradiated in deuterium plasma E =90 eV, Φ = 3,4 · 10 25 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

7 Tungsten : FLAKING Fig.1. Flakes on the W surface irradiated by high-flux (10 22 m -2 s), high-fluence (up to 10 27 m -2 ) and low-energy (38 eV) deuterium plasma at the temperature T= 618 K [1]. The temperature of flaking is specific for cooper irradiated with moderate fluxes (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.

8 Tungsten : MELTING OF FLAKS Local melting of W surface irradiated by high-flux (10 22 m -2 s), high-fluence (up to 10 27 m -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.

9 Tungsten : SUB THRESHOLD SPUTTERING Irradiation parameters: Deuterium ions, E i =5 eV, j=(1-2) 10 21 m -2 s, Φ= (0.5-1)×10 26 m -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

10 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.

11 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.

12 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.

13 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.

14 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.

15 Carbon (CFC, graphites)

16 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.

17 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 m 2 ) could be equal to (2-3)×10 22 T.

18 Carbon : Hydrogen retention in redeposited carbon layer Conditions of redeposition(Н/С) ratio in the deposited films 1. Redeposition in tokamaks JT-60 Т10 JET 0.04 0.06 – 0.3 0.1 – 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

19 Carbon tiles under neutron and plasma irradiation Critical issueToday understanding Chemical sputtering Erosion rate is small Tritium accumulation in deposited carbon 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 under neutron irradiation Dimensional changes is not avoidable, if T 800-900 K

20 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 T800-900 K to provide annealing of neutron induced defects and to prevent dimensional changes

21 Boron carbide coating as a renewable protecting coating for ITER divertor- and first wall tiles

22 Parameters of B 4 C. 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

23 Parameters of B 4 C. HYDROGEN RETENTION Fig.4. Temperature dependence of hydrogen retention for different materials (H +, E i =100 eV, j=5.610 19 m -2 ) [7].. [7] L. Begrambekov, O. Buzhinsky, A. Gordeev, E. Miljaeva, P. Leikin, P. Shigin. Physica scripta, N108 (2004), p.72-75.

24 Parameters of B 4 C. 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/m 2 Time, sec Number of cycles Т max, 0 C MPG-8RGT 2,0 2,3 5555 20 < 300 - < 300 5,0 5,8 2,5 10 430 - < 300 5,0 5,8 10 940 - 790 10,0 11,0 13,0 255255 255255 > 1400 - 880 940 [8] Magnetic fusion energy program, annual report SNL(1989)18.( reconstruction of tables 1 and 2)

25 Parameters of B 4 C. Th e product of B 4 C sputtering D/C ratio in sputtered boron carbide coating D/C ratio in redeposited (H/C/B) layer Comments 4.120.22-0.45Redeposited layer could be removed in cleaning discharge Laboratory experiment

26 Parameters of B 4 C. In situ coating deposition Deposition conditionsPlasma providing total dissociation of the molecules of initial substance Devices used for B 4 C deposition Tokamak T-11M, PISCES-B (T e ~40 eV, n e ~2·10 17 m –3, electron flux~2·10 17 m –2 с –1 ) [9]. Initial substance Non-toxic, non-explosive, and non-hazardous carborane (C 2 B 10 H 12 ) 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., 313316 (2003) 214.

27 Advantages of B 4 C coating and further investigations B 4 C coating will provide low rate erosion of divertor tile surfaces. B 4 C coating will prevent erosion of the tiles as well as penetration of tritium into and trapping in the bulk of the tiles. B 4 C coating can be deposited and renewed during regular ITER discharges B 4 C coating can withstand high thermal fluxes (13.0 MWm -2 ) Erosion of B 4 C 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 B 4 C coating in tokamak conditions is needed

28 Hypothetic C-W-B 4 C tile

29 Scheme of B 4 C-W-C tile 1. Water cooled pedestal, 2. Cooling water, 3. Tungsten plate 4. Tungsten sticks, 5. RGT graphite or CFC, 6. B 4 C coating Temperatures of outer and inter boundaries of the tile under 14 MW/ m 2 irradiation (λ W =120 W/mK, λ B4C =20 W/mK, λ РГТ =200 W/mK) Thickness of the partsTemperature Overall thickness of the tile 10 mm Temperature of plasma facing surface 1400C Thickness of B 4 C coating 50 µm Temperature of B 4 C – С boundary 1360 C Thickness of RGT plate 4 mm Temperature of C – W 900 C Thickness of tungsten sticks 5 mm Temperature of W – pedestal 200 C Thickness of tungsten plate 1.5 mm 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.

30 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-B 4 C 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.

31 Thank you for your attention!

32 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.

33 Introduction High-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 (B 4 C) 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-B 4 C tile is discussed


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