EAST(HT-7U ) Physics and Experimental Plan

Transcription

1 EAST(HT-7U ) Physics and Experimental Plan EAST team, presented by Jiangang Li Institute of Plasma Physics, CAS 1 st EAST IAC meeting Hefei, Oct.10-11

2 The mission of EAST To explore the methods to achieve steady-state operation for tokamak fusion device using noninductive current drive; To integrate optimized plasma performance and continuous operation to establish the scientific and technological bases for next step burning plasma devices.

3 The role of EAST Test bench of ITER for advanced steady state operation. One of the key devices for steady state operation. Supply data base for SST reactor β N *H DIIID AsdexU JT60U JET ITER HT7 HT7U 4 Advanced tokamak 2 normal tokamak t H /τ E

4 Physics objective of EAST Long sustainment of inductive high performance plasma (~ t wall ) at Te>10keV, n e >10 20 m -3 Long sustainment of Full CD hot plasma (1000s) at Ip=1MA. Realization of advanced mode of operation with reasonably high performance (β N >2, H 89 >2) for the pulse length up to 1000 seconds. Long sustainment of high beta (β N ~3) plasma. (high k & d, Mode-control-coil for RWM) Divertor optimization (compatibility with high δ and β, Doped C and metallic PFC, forced cooling Divertor, long particle exhaust) Reduced Activation Ferritic steel, low radioactive SS and ODS W of PFC

5 Three phases of the operation

6 EAST Advanced Tokamak Program

7 Phase I commissioning (< 2 years) Testing for machine capacity Basic ohmic operation for Limiter, SN,DN RF and LHCD coupling RF wall conditioning for ITER Basic diagnostics for machine operation Establish for real-time control system. Ip >1.0MA, n e >1x1020 m -3 Te > 2keV, Ti>1.5keV, t >30s, τ E >50ms.

8 Start-up Reliable ohmic system with NbTi ST coils I (KA) t (s) Ip Vp=8V, V-S db/dt =7T/s (50ms) Iv I(KA) OH Ip(*100) VF Vp(V) V-S(wb) Full data base for difference operation scenarios t (s)

9 Phase I Basic experiments Divertor configuration operation (30-60s Testing the capacities of PF and fast internal feedback coils. Main objectives I Low shaping operation K=1.5 δ=0.35 II RF and LHCD system testing, coupling, impurity suppression. III RF+LHCD, operation window, full wave current drive 30s) and effective heating (>5keV) IV Divertor material and structure testing. PW interaction, doped graphite PWM, heat load for long pulse, particle exhausting.

10 Basic Diagnostics Mainly for machine operation Electro-magnetic coils (for EFIT and control) Te: (ECE,SX-PHA) Ne (FIR interferometer, DCN) FW&Divertor Langmuir probes,visible spectroscopy,visible and IR CCD ; Bolometer impurity monitors HX-ray, H α array; MHD: Magnetic probes, SX array Vacuum Gauge,RGA

11 Phase II experiments 5-8years Ip = 1.0MA t = 400s, SN DN configuration P LHCD = 3-4MW, P RF = 3MW, P BNI =4MW B t =3.5Tesla, q 95 =5.1,ne=3x10 19 cm -3 H 89 =1.5 H 89 =2 H 89 =3 E (ms) <nt>(10 19 m -3 kev) W(MJ) f b (I b / I p ) N

12 Advanced Tokamak Research Thrusts Extend pulse length to steady-state condition with fully non-inductive current drive (LHCD + Bootstrap). Profile Control via Lower Hybrid Current Drive, IBW,MCCD, ICRH at the time scale which is well beyond wall saturate time. Understanding, control and sustainment of Internal Transport Barriers, with Ti. Te and no momentum input. Divertor optimization by the large plasma shaping, good power and particle handling structure. Increase through profile optimization to MHD limit. Advanced PFM and Wall conditioning technique for tokamak reactor. New diagnostics and upgrades

13 Plasma Control Equilibrium, shape and position PF coils+ internal fast feedback coil, Develop state-of-the-art simulations and models for real-time and steady-state condition Real time data collection for SSO profile control Error field :correction coils(br/b~ 5-8x10-5 ) Toroidal field ripple and fast particle losses : The ferromagnetic material :ripple 2.7% to less than 1-1.5%. MHD Instabilities (NTM,RWM) Tailoring J(r) by LHW,IBW, NBI, RFC coils Disruption mitigation: Ar,Ne puff(gp, MBI), killer pellets Profile controls Ne: NBI, ICRF, IBW, pellet, SMB Te(Ti): LHCD, IBW, NBI, ECRH, MCCD, ICRF Transport and turbulence: IBW, LHCD, ECRH ITB: IBW+LHCD, IBW+ECRH

14 Transport and Pressure Profile Control Improve understanding of ITBs in LHCD plasmas and with off axis IBW. What is the threshold condition? Role of rotation? Detailed profiles, time behavior of, D? Barrier location control. What determines the barrier location (usually r/a~ )? What is the role of zonal flow Can bootstrap current be expanded for more attractive AT scenario? Current scan, fast current ramps. Improving performance ( β N, H 89 ). Strong shaping, κ~1.8, δ~ How can we maximize energy confinement, bootstrap current? Does regime extend to higher T, lower *? Impurity contribution for SSO state

15 Transport and Pressure Profile Control Investigate influence of magnetic shear on ITB formation, location and transport profiles. Use LHCD +IBW and compared with LHCD+ MCCD to control j(r). Using LHCD+ECRH and compared with LHCD+IBW Study effect of flow drive on barriers. Depends on DLIBW/MCIBW flow drive tests. Testing the IBWCD at high density Can it be an active barrier control tool? Adjust heating profile to modify Te(R) Two ICRF frequencies(high harmonic IBW heating). Two frequency LH Heating Optimize density, temperature, bootstrap profiles for compatibility with LHCD, maximum non-inductive CD scenario.

16 Inductive hybrid operation Mode-I: Long pulse Ip=1.0MA, n e =2x10 19 m -3, P IBW =1.5MW, B T =3.5T, P LHCD =2.0MW, P NBI =2MW, Wj=1.0MJ, I bs /Ip=28%, I LHCD =75%, τ E =128ms, Te=8keV, t=400s, β N H 89 >3. Mode-II: high density Ip=0.4MA, n e =5x10 19 m -3, P IBW =1.5MW, B T =2T, P LHCD =2.0MW, P NBI =4MW, Wj=0.5MJ, I bs /Ip=58%, I LHCH =10%, Ti=3keV, τ E =58ms, Te=5keV, t=60s, β N H 89 >4, t H /τ E > 1000, ne/n GW =0.8.

17 Steady-state H-mode operation (for ITER) weak central shear Ip=0.6MA, B T =2.4T, q 95 =6.3, n e /n GW =0.5 H 89 =2.3, N=3.2 Vp=0 f bs =46%, f CD =54%, t=400s. P IBW =1.5MW, P NBI =4MW P LHCD =2.0MW.

18 Advanced diagnostics for physics understanding FIR Polarimetry MW Reflecmeter. TS DNB MSE MW Image CO 2 scattering Lp Array CXRS High resolution ECE Material Station Mat. St.

19 EAST Divertor B2-EIRENE Simulation shows acceptable SN and DN configurations.

20 For density >2x10 19 m -3, the power load and edge condition is fine up to 10MW injection power

21 FM:2-3MW/m 2,doped C with SiC coating, Bolted type heat sink. Divertor legs: 8-10MW/m 2, W coating on high performance C, C brazed to Cu heat sink Poloidal limiters Divertor cassette modules, easy to change and alignment. Internal fast feedback coils for VDI control Using internal crypump in later phase. Structures of Divertor and internal components

22 FW material developing stratagem Evaluate the performance of different materials for divertor/first-wall FWM: < 2MW/m 2, technically ready High-field side: SiC coating on the doped graphite, bolted heat sink; Low-field side: W coating on the Ferritic steel Divertor: 4-6MW/m 2 (8-12MW/m2) SiC/B 4 C coating on the high performance doped graphite ( inner leg), C brazed to Cu heat sink. W coating (0.5mm) on the high performance graphite(out leg), C brazed to Cu heat sink. W coating (1-2mm) on the Cu heat sink. W coating (mm) on low radioactive steal (CLAN, similar with H82), high Tw operation for T inventory.

23 Plasma wall interaction in Long Pulse AT operation To understand the flows and exchanges of fuel and impurity particles between the plasma and the facing materials for fuel and impurity control. RF wall conditioning technique in divertor device for ITER( cleaning, isotopic control, boronization, siliconization). Develop RF Tritium removal techniques that are applicable for ITER. Steady-state erosion and redeposition. In-situ control of T-codeposition and migration by surface temperature control. Life time of graphite and W under the SSAT base.

24 Power and particle handling in Long Pulse AT operation Steady-state power load: Obtain the radiative divertor (strong D 2 gas puffing, SMB puffing). "Puff and Pump" and injected impurity radiates in divertor. Good heat load removing structure ( C brazed to Cu heat sink). Pulsed power load: Mitigated disruption with noble gas injection and killer pellet. DN operation allows precise control of heat and particle fluxes. ELM-free operation eliminates pulses divertor heat and particle loads. ELM-free QH-mode edge is favorable for sustained ITB operation by wave only( experienced in HT-7), which could be important on EAST with limited heating power. Magnetic balance controls to Elm and H-mode power threshold and particle exhaust rate ( to find a way to avoid the type-i Elm).

25 Main difficulties of EAST Very limit budget Very few good young scientists Very little Codes and simulations Little experiences in NBI and ECRH Very few advanced diagnostics Need further international cooperation

26 Challenges and opportunities Challenges : Superconducting techniques. Continuous heating and current driven techniques. Steady-state operation and real-time control. Heat and particle removal in Steady-state condition. Neutron shielding Opportunities: Evolution of the current profile on the time scale much longer than the resistive time in fusion-relevant tokamak plasma. Behavior of the transport barriers in steady-state operation mode MHD stability on the steady-state base. Test bench for ITER. A key device for steady-state operation in the fusion community.

27 Summary EAST will try to contribute to the techniques for the steady state high performance plasma operation. It will extend the advanced tokamak research to the steady state operation condition and supply an useful data base for next step superconducting tokamak. It can be a world-wide fusion collabratory. The budget and personnel is limited. It needs more efforts on codes and simulations. Wide international cooperation and helps are welcomed.

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