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L.B. Begrambekov 1 , O.I. Buzhinsky 2 , A.M. Zakharov 1

19 th International Conference on Plasma Surface Interaction in controlled fusion devices San Diego, California. May 24-28, 2010. Possibility of protecting renewable B 4 C coating application in ITER. L.B. Begrambekov 1 , O.I. Buzhinsky 2 , A.M. Zakharov 1.

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L.B. Begrambekov 1 , O.I. Buzhinsky 2 , A.M. Zakharov 1

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  1. 19th International Conference on Plasma Surface Interaction in controlled fusion devices San Diego, California. May 24-28, 2010 Possibility of protecting renewable B4C coating application in ITER. L.B. Begrambekov1, O.I. Buzhinsky2, A.M. Zakharov1 1National Research Nuclear University MEPhI, 31 Kashirskoe sh., Moscow 115409, R.F 2Troitsk Institute for Innovations and Fusion Research, Troitsk, 142190, R.F.

  2. OUTLINE • Introduction • Problems of tungsten divertor tiles • Problems of carbon (CFC, graphites) divertor tiles • B4C as a renewable protecting coating for the tiles. • Hypothetic C-W-B4C tile

  3. Tungsten

  4. Tungsten . BLISTERING(cones and pyramids) The temperature of blistering is specific for cooper irradiated with moderate fluxes. (T=480-520 K) 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, [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+ , Ei =400 eV, j= 1,2×1020m-2s, Φ= 5×1022m-2, T=1150 K

  6. Tungsten : BLISTERING Рис. Blisters on the W surface irradiated in deuterium plasma E =90 eV, Φ = 3,4 · 1025м-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 The temperature of flaking is specific for cooper irradiated with moderate fluxes (T=618 K) 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]. [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 The temperature of flake melting Is estimated to be (T=1300-1400 K). It is the temperature range of cooper melting point 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 [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, 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 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 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 [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

  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. J. Roth, R. Preuss, W. Bohmeyer, S. Brezinsek, et al. Nucl. Fusion 44 (2004) L21. 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] [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 m2) could be equal to (2-3)×1022 T.

  18. Carbon:Hydrogen retention in redeposited carbon layer

  19. Carbon tiles under neutron and plasma irradiation

  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 T≥800-900 K to provide annealing of neutron induced defects and to prevent dimensional changes

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

  22. Parameters ofB4C.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 ofB4C. 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. Physicascripta, N108 (2004), p.72-75.

  24. Parameters ofB4C.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]. [8] Magnetic fusion energy program, annual report SNL(1989)18.( reconstruction of tables 1 and 2)

  25. Parameters ofB4C.The product of B4C sputtering Laboratory experiment

  26. Parameters ofB4C.In situ coating deposition [9] Buzhinskij O.I., Otroschenko V.G., Whyte D.G. et al. J. Nucl. Mater., 313—316 (2003) 214.

  27. Advantages of B4Ccoating and further investigations • ● B4C coating will provide low rate erosion of divertor tile surfaces. • ● B4Ccoating will prevent erosion of the tiles as well as penetration of tritium into and trapping in the bulk of the tiles. • ● B4Ccoating can be deposited and renewed during regular ITER discharges • ●B4Ccoating can withstand high thermal fluxes (13.0 M∙W∙m-2) • ● Erosion of B4Ccoating 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 B4Ccoating in tokamak conditions is needed

  28. Hypothetic C-W-B4C tile

  29. Scheme of B4C-W-C tile 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. 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 under 14 MW/ m2 irradiation (λW=120 W/mK, λB4C=20 W/mK, λРГТ=200 W/mK)

  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-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.

  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 (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

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