18 th International Conference on Plasma Surface Interaction Toledo, May 26-30, 2008 Plasma-wall interactions and plasma behavior in fusion devices with liquid lithium plasma facing components S.Mirnov TRINITI, Troitsk, Moscow reg. Russia

Outline 1.Lithium as a new paradigm of fusion reactor PFC 2. Experimental basis 3. Liquid lithium in Capillary Porous System (CPS) 3.1 Lithium CPS properties 3.2 Li CPS cooling 3.3 Li CPS resistance to disruptions 3.4. Lithium radiation. 4 The liquid lithium compatibility with tokamak plasma 4.1 T-11M and FTU experiments 4.2. Lithium radiation in T-11M and FTU 4.3. The lithium screening 4.4 The liquid lithium erosion 4.5 Deuterium retention and removal 5.ITER experiment. Conclusions

1. Lithium as a new paradigm of fusion reactor PFC

The call of energetic crises: what can give fusion to energetic ? 1. DEMO with W-wall at 2050Y? 2. Steady state Volumetric Neutron Source (VNS) with 14 MeV neutron load MW/m for U production from Th or for actinides (Am, Np) transmutation ? But w e have not solution for steady state PFC

Melting, cracking, blistering, abnormal ion sputtering, low-cycles fatigue of solid PFC and dust problem in tokamak conditions look as a main obstacles for steady state tokamak-reactor and VNS. As one of solution seems the emitter-collector concept of liquid metal PFC

COOLING Tokamak first wall Non coronal radiation emitter LM Collector Scheme of “emitter – collector” tokamak liquid metal (LM) limiter

The main request to LM in emitter-collector scheme is the poor its penetration to the hot plasma center, but “impurity penetration increase with increase Z” (for example, R-02 W.Fundamenski). Li have the lowest Z (from 1 to 3) Why 1?

Li properties Electric conductivity (ohm cm) , Specific weight (g/cm 3 ) - 0,5 Heat conductivity (W/cm grade) - 0,53 (627 0 С) Heat capacity (kal/cm 3 grade) - 0,5 Melting temperature - 180,5 0 С Evaporation temperature С Evaporation heat 1,5 (eV/at) First ionization potential 5.6 eV Second ionization potential 75 eV (!) Third ionization potential 122 eV Total “ionization cost” – eV ”Radiation cost” of Li +++ ionized by electrons with Т е =30eV 1200 (eV/at)

Li+ fraction as function of lithium confinement time , plasma density (in cm-3.) and electron temperature Te.

The main problems of liquid lithium use in tokamak praxis: 1) the liquid metal splashing under the JxB forces during MHD instabilities and disruptions, 2) the possible anomalous lithium erosion as a result of plasma-liquid lithium interaction, 3) the problem of heat removal as a prevention of strong lithium evaporation, 4) the problem of the tritium removal from lithium.

2. Experimental basis

Test of lithium compatibility with tokamak plasma The first wall lithiation experiments: TFTR (1996), T-11M( ), CDX-U( ), FTU(2005), T-10 (2006), TJ-II (2008) The experiments with liquid lithium limiters: T-11M, CDX-U, FTU The future liquid lithium experiments: FTU(2010?), NSTX(2009), LTX(2010?), T-15(2012?)

The main parameters of LLL tokamaks TOKAMAKST-11MCDXU FTU R/a [cm/cm] 70/20, k=1 34/22, k=1.6 84/20, k=1 B T [ T ] J p [ kA] 100 < ∆t [ms] 250 < [10 19 m- 3 ] T e [eV] Kind of Li limiterCPS free Li Surface “tray lim.” CPS

Li-CPS limiter experiment in T-11M (TRINITI-”Red Star”) V.A.Evtikhin et al. Plasma Phys. and Controlled Fusion 44 (2002) 955 Scheme of T-11M experiment T-11M parameters: R/a=0.7/0.2m, Bt=1T, Jp= kA, Plim ≈ 10-20MW/m2, δt= s SXR

Scheme of FTU experiment G. Mazzitelli e.a. 21th IAEA Conf. on Fusion Energy Chengdu, (2006) EX/P4-16.

3. Liquid lithium in Capillary Porous System (CPS)

Li capillary pore structure (CPS) The idea to use LM in tokamaks as PFC was advanced basing on the surface tension forces in capillary channels for compensation of ponder- motive forces. These capillary channels ( microns) may be realized in the form of so called capillary-pore systems (CPS) (V.A.Evtikhin et al.1995). Self-regeneration of liquid metal surface, contacted with plasma is an intrinsic property of such structures. Mo-mesh with lithium filling and without it - CPS as PFC element

Heat removal problem Thin (1-0.6mm) Li CPS-limiter experiment Scheme of Li CPS-Mo limiter The ends of thin (Mo,W) Li CPS (δ<1mm) should sink in Li-reservoir so that it works as steady-state Li-wick. The main heat flux should go across thin CPS layer to Mo cooled backing. Fresh lithium can flow to limiter surface along wick. Cooling – water (experimental tokamak), eutectic Pb+Mg,K ?

Resistance of Li CPS to tokamak disruption (V.A. Evtikhin et al. Plasma Phys.&Contr. Fus. 44 (2002) 995) In the plasma gun experiments (q=4-5MJ/m2 t= ms) it was shown, that main part of the plasma energy (~97- 99%) was radiated during shot in CPS front lithium cloud, which plays the role of a shielding layer. This result has been confirmed in T-11M: only 30-50J of about 0.7 kJ of total plasma energy loss has been found to reach the rail limiter during disruptions. The solid basis of CPS limiter (Mo, SS) had no damages after more than 2000 of plasma shots with 5-10% disruptions.

Lithium radiation (non coronal) V.B. Lazarev V B et al. 26th EPS Conf.on Contr. Fus.&Plasma Phys. ECA 231(1999) 845

Lithium radiation in non-coronal regime

T-11M, FTU experiments

Plasma-rail limiter interaction T-11M Cold exposition initial T lim 200C

FTU experiment. Li CPS limiter after plasma exposition No Surface Damage

T-11M, FTU- plasma radiation in shots with “high” level of Li injection

Space distribution of plasma radiation for Li (blue) and C(red) limiters (Т-11М). The non-coronal lithium radiation can remove up to 80% P OH power to vessel wall. a, m

FTU visible light emission in shots with low and high Li injection

FTU-total radiation with/witout Li injection

Lithium screening (localization close plasma boundary, Z eff (0)≈1) as main surprise of lithium tokamak experiment

D.K. Mansfield et al, Phys. Plasmas 3 (1996) 1892

FTU experiment (FAC2006). Zeff behaviour during all the experimental campaign After lithium limiter insertion Shots Zeff

Liquid lithium erosion (T-11M from LiI behavior) S.V.Mirnov, E.A.Azizov, V.A.Evtikhin, et al. Plasma Phys. and Contr. Fus. 48 (2006) 821.

Lithium erosion versus limiter temperature in T-11M Lithium sputtering versus target temperature (J.P.Allain et al. J.Nucl.Mat )

The behaviour of both curves (LiI and Sput.Y) versus T can be approximated by exponential functions ~ exp – E k /T ( k=I,S ). In temperature interval C the characteristically lithium sputtering energy Es is equal 0.22 ± 0.02eV and lithium emission energy E I is equal 0.2 ± 0.02eV. This practically equality permits us to assume, that real mechanism of liquid lithium erosion in tokamak limiter is the same, as liquid Li erosion by simple ion bombardment in beam experiment.

Deuterium and Helium retention in Li CPS (T-11M) The Li-limiter heating to 450 o C shown a desorbtion of the captured deuterium at temperatures higher than 320 o C. Helium was sorbed by lithium films, too. But heating of chamber wall to C prevented its retention. (V.A.Evtikhin et al. Plasma Phys. and Controlled Fusion 44 (2002) 955)

Deuterium removal from T-11M LL limiter and LL targets M.J.Baldwin, R.P.Doerner, S.C.Luckhard and R.W.Conn, Nucl.Fusion 42 (2002) Y.J. Furuyama, J Nucl. Mat (2003) 288.

The approximation of this curves by exponential function ~ exp – E k /T again gave E 0R = eV - the characteristically deuterium removal energy. It is close to E 0VP = eV- energy of lithium evaporation, but it is very far from the characteristically energy of lithium hydrides decomposition (≈ 2eV), which decayed at temperatures higher than 600C. Therefore, it can be conclude that main part of retained deuterium wasn’t captured by lithium in the form of hydrides (deuterides), but it was just dissolved in lithium and removed, probably together with its evaporation. Lithium heating up to C seems to be sufficient to remove from lithium almost all deuterium and, probably, tritium also.

Li emitter-collector experiment for ITER

Li mushroom limiter (ML) position in ITER equatorial port

Conclusions The tokamak experiments with liquid Li limiters have shown good compatibility of liquid lithium PFC with tokamak plasma. Main properties of liquid lithium PFC on CPS basis in tokamak plasma can be summarized as followed: 1.The surface tension forces in CPS may be successfully used to solve the problem of liquid lithium splashing during MHD-events 2. The surface tension forces may afford the PFC regeneration in steady state devices.

3. The experiments with hydrogen (deuterium) and helium plasmas on T-11M tokamak with Li – CPS limiter have shown: -no serious spontaneous lithium ejections under a heat flux to the limiter up to the power load of MW/m2 and lithium temperature lover 600 grad C have been observed. -a total lithium erosion of Li PFC during the interaction with tokamak SOL plasma is close to the level of hydrogen and lithium ion sputtering. - the lithium non-coronal radiation protected the limiter from high power load during disruptions. - In ordinary tokamak regimes the Li ions, circulated in limiter SOL, allow the removal up to 80% P OH power to vessel wall by non- coronal lithium radiation.

-the solid basis of CPS limiter had no damages after more than 2000 of plasma shots with 5-10% of disruptions -the temperature of hydrogen isotopes removal from Li after plasma bombardment is C (for helium C). Therefore, at high PFC temperatures ( C) tritium capture can be minimized - the separation of helium and hydrogen isotopes is possible in lithium circuit with lower PFC temperatures The lithium ion behaviour in “lithium tokamaks” (TFTR, T-11M, CDX-U, FTU, NSTX) permits us to believe the existence of lithium screen mechanism. It should be tested in ITER-like tokamaks and in stellarators too.

4. But for successful use of lithium PFC in future steady state tokamaks should be investigated: - problem of lithium cooling in steady state mode, - the lithium self-sputtering and lithium sputtering by hydrogen isotopes during long term PFC exposition in tokamak plasma (like TRIAM), - physical origin of lithium screening effect.

5. The current experiments gave hope that liquid lithium PFC can be used for steady state tokamak reactor like ITER and DEMO. The following problems of such reactor might be solved: wall and divertor plates erosion (by self-recovery), “dust” accumulation and redeposition, tritium recovery, radiative cooling core plasma due to the high Z impurity, heat removal during steady state regime and disruption. The ITER Li-limiter experiment with combined Li-emitter and collector, which could decrease the power load to divertor plate, can be suggested today.

Conclusion of optimist: Li is way to future fusion. Conclusion of pessimist: Li is direction to future fusion

Thank you for attention