Y. Ueda, M. Fukumoto, H. Kashiwagi, Y. Ohtsuka (Osaka University)

Slides:



Advertisements
Similar presentations
J. Roth, EU PWI TF, SEWG Fuel Retention, Cadarache, June 15, 09 Tritium inventory: Joint international scaling for ITER WP09-PWI-01-01/IPP/PS Status by.
Advertisements

Report on SEWG mixed materials EU PWI TF meeting Madrid 2007 V. Philipps on behalf of SEWG members Mixed material formation is a among the critical ITER.
Max-Planck-Institut für Plasmaphysik EURATOM Assoziation Interaction of nitrogen plasmas with tungsten Klaus Schmid, A. Manhard, Ch. Linsmeier, A. Wiltner,
PWI Modelling Meeting – EFDA C. J. OrtizCulham, Sept. 7 th - 8 th, /8 Defect formation and evolution in W under irradiation Christophe J. Ortiz Laboratorio.
Kazuyoshi Sugiyama, SEWG meeting, Culham, July Outline: 1.Introduction 2.Experimental procedure 3.Result 4.Summary Kazuyoshi Sugiyama First.
SEWG Gas Balance 2007 © Matej Mayer First results on deuterium depth profiling in W tiles M. Mayer 1, V.Kh. Alimov, V. Rohde 1, J. Roth 1, A. Herrmann.
Max-Planck-Institut für Plasmaphysik EURATOM Assoziation K. Schmid SEWG meeting on mixed materials Parameter studies for the Be-W interaction Klaus Schmid.
Member of the Helmholtz Association Carbon based materials: fuel retention and erosion under ITER-like mixed species plasma conditions Arkadi Kreter et.
Institute for Plasma Physics Rijnhuizen D retention in W and mixed systems in Pilot-PSI G. De Temmerman a, K. Bystrov a, L. Marot b, M. Mayer c, J.J. Zielinski.
D retention in O-covered and pure beryllium
H and He cluster formation in W Krister Henriksson Accelerator laboratory.
Introduction to Plasma-Surface Interactions Lecture 6 Divertors.
18 th International Conference on Plasma Surface Interaction in Controlled Fusion Toledo, Spain, May 26 – 30, Deuterium trapping in tungsten damaged.
© Olga Ogorodnikova, 2008, Salamanka, Spain Current status of assessment of Tritium inventory in all-W device O.V. Ogorodnikova and E. d’Agata.
ADSORPTION STATES OF PROTIUM AND DEUTERIUM IN POLYMER HYDROCARBON FILMS FROM T-10 TOKAMAK V.G. Stankevich 1, N.Yu. Svechnikov 1, L.P.Sukhanov 1, K.A.Menshikov.
1 EFFECTS OF CARBON REDEPOSITION ON TUNGSTEN UNDER HIGH-FLUX, LOW ENERGY Ar ION IRRADITAION AT ELEVATED TEMPERATURE Lithuanian Energy Institute, Lithuania.
L.B. Begrambekov Plasma Physics Department, Moscow Engineering and Physics Institute, Moscow, Russia Peculiarities, Sources and Driving Forces of.
PISCES R. Doerner, ITPA SOL/DIV meeting, Avila, Jan. 7-10, 2008 Mixed plasma species effects on Tungsten M.J. Baldwin, R.P. Doerner, D. Nishijima University.
Dynamic hydrogen isotope behavior and its helium irradiation effect in SiC Yasuhisa Oya and Satoru Tanaka The University of Tokyo.
Japan-US Workshop held at San Diego on April 6-7, 2002 How can we keep structural integrity of the first wall having micro cracks? R. Kurihara JAERI-Naka.
Japan PFC/divertor concepts for power plants. T retention and permeation  Problems of T retention would not be serious…. Wall temperature will exceeds.
1 R. Doerner, ARIES HHF Workshop, Dec.11, 2008 PMI issues beyond ITER Presented by R. Doerner University of California in San Diego Special thanks to J.
K.Umstadter –-Laser+D on W PISCES Effects of transient heating events on W PFCs in a steady-state divertor-plasma environment Karl R. Umstadter, R. Doerner,
Deuterium retention mechanisms in beryllium M. Reinelt, Ch. Linsmeier Max-Planck-Institut für Plasmaphysik EURATOM Association, Garching b. München, Germany.
Dynamic evolution of mixed materials bombarded with multiple ions beams and impurities Tatyana Sizyuk Ahmed Hassanein School of Nuclear Engineering Center.
Ion-Driven Permeation of Deuterium through Tungsten Motivation Permeation experiment Results Next steps A. V. Golubeva, M. Mayer, J. Roth.
R. Doerner, May 9, 2005 PFC Program Review, PPPL PISCES ITER-simulation experiments on Mixed-Materials (Be, C, W) R. P. Doerner, M. J. Baldwin and D. Nishijima.
R. Doerner, IAEA CRP on H in Materials, Vienna, Sept. 26, 2006 Mixed-material studies in PISCES-B R. P. Doerner, M. J. Baldwin, J. Hanna and D. Nishijima.
Salamanca.ppt, © Thomas Schwarz-Selinger, 03. Juni 2008 G. S. Oehrlein*, T. Schwarz-Selinger, K. Schmid, M. Schlüter and W. Jacob Interaction of Deuterium.
Measurement and modeling of hydrogenic retention in molybdenum with the DIONISOS experiment G.M. Wright University of Wisconsin-Madison, FOM – Institute.
Sachiko Suzuki 1, Akira Yoshikawa 1, Hirotada Ishikawa 1, Yohei Kikuchi 1, Yuji Inagaki 1, Naoko Ashikawa 2, Akio Sagara 2, Naoaki Yoshida 3, Yasuhisa.
K. Sugiyama, 9th International Workshop on Hydrogen Isotopes in Fusion Reactor Materials, Salamanca, June 2-3, Max-Planck-Institut für Plasmaphysik.
PISCES The effects of high fluence mixed-species (D, He, Be) plasma interactions with W 18 th Int. Conf. on Plasma Surface Interactions, May 26–30, (2008)
Tritium Retention in Graphite and Carbon Composites Sandia National Laboratories Rion Causey Sandia National Laboratories Livermore, CA
R. P. Doerner, 2 nd PMIF Meeting, Juelich, Sept , 2011 Plasma interactions with Be surfaces R. P. Doerner, D. Nishijima, T. Schwarz-Selinger and.
R. Doerner, ITPA SOL/DIV meeting, Avila, Jan. 7-10, R. P. Doerner, G. De Temmerman, M.J. Baldwin, D. Nishijima Center for Energy Research, University.
第16回 若手科学者によるプラズマ研究会 JAEA
Tritium burnup fraction C. Kessel, PPPL ARIES Project meeting, July 27-28, 2011 Gaithersburg, MD.
Depth-profiling and thermal desorption of hydrogen isotopes for plasma facing carbon tiles in JT-60U (Long term hydrogen retention) T. Tanabe, Kyushu University.
*This work was supported by the United States Department of Energy
Olga Ogorodnikova, 2008, Salamanka, Spain Comments to modelling of hydrogen retention and permeation in tungsten O.V. Ogorodnikova Max-Planck-Institut.
Introduction to Plasma- Surface Interactions Lecture 3 Atomic and Molecular Processes.
Transport of deuterium - tritium neutrals in ITER divertor M. Z. Tokar and V.Kotov Plasma and neutral gas in ITER divertor will be mixed of deuterium and.
Effects of tungsten surface condition on carbon deposition
Plasma-wall interactions during high density operation in LHD
PSI 2008 Toledo May 2008 © Matej Mayer Carbon balance and deuterium inventory from a carbon dominated to a full tungsten ASDEX Upgrade M. Mayer a, V. Rohde.
Compositional dependence of damage buildup in Ar-ion bombarded AlGaN K. Pągowska 1, R. Ratajczak 1, A. Stonert 1, L. Nowicki 1 and A. Turos 1,2 1 Soltan.
O AK R IDGE N ATIONAL L ABORATORY U. S. D EPARTMENT OF E NERGY 1 Update on Helium Retention Behavior in Tungsten D. Forsythe, 1 S. Gidcumb, 1 S. Gilliam,
Edge-SOL Plasma Transport Simulation for the KSTAR
1 US PFC Meeting, UCLA, August 3-6, 2010 DIONISOS: Upgrading to the high temperature regime G.M. Wright, K. Woller, R. Sullivan, H. Barnard, P. Stahle,
R. Doerner, ITPA SOL/DIV meeting, Avila, Jan. 7-10, Be deposition on ITER first mirrors: layer morphology and influence on mirror reflectivity G.
The effect of displacement damage on deuterium retention in plasma-exposed tungsten W.R.Wampler, Sandia National Laboratories, Albuquerque, NM R. Doerner.
R. Doerner, PFC Program Meeting, MIT, July 8-10, 2009 Mixed Interactions of W, Be, C, D & He R. Doerner for the PISCES Team In collaboration with members.
1 Deuterium retention and release in tungsten co- deposited layers G. De Temmerman a,b, and R.P. Doerner a a Center for Energy Research, University of.
1 Russian Research Center” Kurchatov Institute” Alexander Ryazanov Charge State Effects of Radiation Damage on Microstructure Evolution in Dielectric Materials.
MOLIBDENUM MIRRORS WITH COLUMN NANOGRAIN REFLECTING COATING AND EFFECT OF ION- STIMULATED DIFFUSION BLISTERRING RRC «Кurchatov Institute» А.V. Rogov, К.Yu.Vukolov.
Application of optical techniques for in situ surface analysis of carbon based materials T. Tanabe, Kyushu University Necessity of development of (1) in-situ.
10th ITPA conference, Avila, 7-10 Jan Changes of Deuterium Retention Properties on Metals due to the Helium Irradiation or Impurity Deposition M.Tokitani.
9 th International Workshop on Hydrogen Isotopes in Fusion Reactor Materials Salamanca, Spain, June 2 - 3, Simulation experiments on neutron damage.
PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION Carbon as a flow-through, consumable PFC material Peter Stangeby University of.
Radiation Damage Quick Study Edward Cazalas 3/27/13.
J. Roth: ITPA SOL/DIV, Avila, Jan Prediction of ITER T retention levels with W PFCs J. Roth, and the SEWG Fuel retention of the EU Task Force on.
Member of the Helmholtz Association Fuel Retention and Erosion of Metallic Plasma-Facing Materials under the Influence of Plasma Impurities A. Kreter 1,
Investigation of the Performance of Different Types of Zirconium Microstructures under Extreme Irradiation Conditions E.M. Acosta, O. El-Atwani Center.
Effects of Pulsed He + Irradiation on Tungsten Surfaces R.F. Radel, G.L. Kulcinski, S.J. Zenobia HAPL Meeting-ORNL March 22 nd, 2006 Fusion Technology.
Tatyana Sizyuk Ahmed Hassanein School of Nuclear Engineering
Presented by T. Sugie (ITER-IT) N. Yoshida (Kyushu University, Japan)
ITERに係わる原子分子過程 Atomic and Molecular Processes in ITER SHIMADA, Michiya ITER International Team Annual Meeting of Japan Society of Plasma Science and Nuclear.
Co-Al 시스템의 비대칭적 혼합거동에 관한 이론 및 실험적 고찰
Deuterium retention for sample temperature of 500 K
Presentation transcript:

Effects of simultaneous impurity ion irradiation on tritium behavior near tungsten surface Y. Ueda, M. Fukumoto, H. Kashiwagi, Y. Ohtsuka (Osaka University) R. Akiyoshi, H. Iwakiri, N. Yoshida (Kyushu University) 9th International Workshop on Hydrogen Isotopes in Fusion Reactor Materials June 2 -3, 2008 Salamanca, Spain Osaka University

Surface phenomena affecting T behavior Deposition layer Trapping site for T Diffusion barrier for T Mixing layer Desorption barrier He bubble layer Radiation damage by n w Ne, Ar D T O He C, Be mixing layer (collision mixing) T deposition layer T Erosion   T T T He bubbles mixing layer (diffusion mixing) T T W Diffusion barrier  Radiation damage

Steady-State High-Flux Dual Ion Beam Flux:~1020 m-2, Energy: 0.15~3 keV Blanket first wall condition Osaka University

Enhancement of blister formation by carbon impurity C concentration in H beam increases C layer W Beam irradiation area Carbon deposition (no blisters) Formation of blisters No blisters Small amount of carbon (less than 1%) in ion beam can enhance blister formation on W. Experimental conditions Beam Energy: 1keV H3+, Flux : (3-4)x1020 Hm-2s-1 Temperature : 653 K Sample : pure W with mirror polished Osaka University

Mechanism for blistering Implantation of H (a few nm ~ 20 nm) grain ejection Accumulation of H at grain boundaries Dome-like blisters > 1 µm H Cross section of blister (K-dope W) Osaka University

W and C mixing layer reduced desorption Atomic composition in tungsten C depth distribution broader than ion implantation range Due to recoil implantation by H High C (~0.9% in the beam) case WC layer reduced recombination of H Enhance bulk diffusion of H Enhance blister formation Low C (~0.1% in the beam) case Low surface C concentration no significant reduction of recombination 1 keV H C: ~0.9% W Blistering C O W 1 keV H C: ~0.1% no Blistering C O Osaka University

E.g. K Tokunaga et al. J. Nucl. Mater. (2004) 337–339, 887. From 300-700 K, thin and thick layers of Be suppresses blister formation. M. Baldwin et al. PSI 18(2008) Blistering & exfoliation of blister caps is a concern for certain varieties of W. Increased retention is associated with the trapping of hydrogen in blisters. E.g. K Tokunaga et al. J. Nucl. Mater. (2004) 337–339, 887. At 550 K a blistered surface is prevalent after exposure to D2 plasma. A thin layer of Be as little as a few 10’s of nm, or thicker, is found to suppress blister formation. D+ ion fluence ~1x1026 m-2

Blister formation under H&He irradiation Small amount of He affected blistering He : ~0.1% has strong effects Suppression of blisters at T>653 K 0.1% He did not change surface mixing layer much. Energy :1 keV (H3+, H2+ , H+) Carbon  :~0.8% Fluence :~7.5 x 1024 m-2 753 K 500 µm 500 µm 653 K 500 µm 500 µm 473 K 20 µm 20 µm 20 µm He : 0.1% He : 0% Osaka University

He bubble could affect H diffusion 1 keV He has slightly longer range than1 keV H (mixed). He bubbles could be formed around the end of ion ranges. He bubbles in W and C mixed layer. He bubbles could be a diffusion barrier for H into the bulk. Stress field affects diffusion? Ion range Osaka University

Flux dependence of blistering C: 0.85%, He:none T = 653 K Flux dependence of blister formation Blistering still appeared by reducing the flux by about 3 ( (2.10.8) x 1020 /m2s ). The number density of blisters decreased. Surface mixing layers (WC) were similar for these cases and formed in the early stage of ion irradiation. He effects on effective flux reduction Since addition of 0.1% He+ to H ion beam completely suppressed blistering, He irradiation corresponded to the case with the flux, lower by more than a factor of 3. High 2.1×1020 /m2s 500 µm 1.3×1020 /m2s 500 µm Flux 0.8×1020 /m2s 500 µm Low

TEM observation of He bubbles He:1.0%, ~2 nm He bubbles He:0.1%, 1~2 nm He bubbles He fluence : 4.1 x 1021m-2. From erosion depth (~300 nm) and ion range (~10 nm), effective He fluence was ~1020 m-2. Only this fluence affected hydrogen diffusion Bubble size and bubble number density had weak dependence on He% and C%. He bubbles were formed in WC layer for C:~0.8%. T = 653 K Fluence : 4.1 x 1024 m-2 TEM observation of near surface structure

swelling rate estimation swelling rate vs. temperature He bubble volume (swelling rate) swelling rate estimation swelling rate vs. temperature 6 5 4 3 2 1 Swelling rate (%) 1100 1000 900 800 700 600 Temperature (K) ━ He 0.1% He 1% Sample 20nm Swelling rate = He bubble volume / total volume Hydrogen diffusion greatly suppressed by only 2% He bubbles.

Effects of He energy on blistering (a) no He Main Ion Beam(1.5 keV : H+C:0.8%) (a) no He ion beam  Blistering (b) 2nd He beam :0.05%(0.6 keV) Blistering (c) 2nd He beam :0.05%(1.0 keV)*    2nd He beam :0.05%(1.5 keV) *    *angle of incidence ~ 40 deg no Blistering (b) He:0.6 keV (c) He:1.0 keV Blistering (0.6 keV He) No blistering (1keV He, 1.5keV He) Osaka University Ion range in tungsten

He effects in ITER (tungsten FW) Energy of ions CX neutrals have relatively high energy (D,T、~600eV) with the flux of mid 1019 m-2s-1. Fuel ions (D,T) have relatively low energy (~200 eV , ~3kTe+2kTi) with the flux of 1020 m-2s-1 . He ions have energy (~300 eV, 3ZkTe +2kTi) with the flux of ~1018 m-2s-1 (R. Behrisch et al., JNM 313-316 (2003) 388.) Ion ranges (normal incidence) CX neutral (T) 8.4 nm (600 eV) T ions  4.2 nm (200 eV) He ions  3.1 nm (300 eV) He implantation may enhance inward diffusion of T and D from CX. Ranges of He and T ions are comparable. He effects? Edge Ti & Te CX Neutrals

He effects in fusion reactors (divertor): ITER Divertor plates He ranges are shorter than T. Te, Ti = 15 eV, He2+ He bubbles are also desorption barrier? Enhancement of T retention? H bubble size ~ range He bubble could not be important. Normal incidence 0.5 m Ion ranges for the edge plasma condition (Temp.~15 eV) Te & ne profiles near divertor SP

Summary and conclusion Simultaneous irradiation of impurity ions (C, He, (Be)) significantly affects hydrogen behavior in tungsten. Surface mixing layer affects hydrogen-isotope behavior Its effects are determined as a balance between reduction of surface recombination and reduction of diffusion into the bulk. He bubble layer can be a diffusion barrier Stress field could reduce diffusion? He effects strongly depend on energy. He ion range  H ion range  Reduction of bulk diffusion He ion range < H ion range  Enhancement of bulk diffusion More study is needed under edge plasma conditions (He ion energy less than ~300 eV). No displacement damage and short ion ranges compared with out experiment. This effect should be properly evaluated and included in T retention estimation in W.