Capability requirements for a PMI research thrust fusion facility

Transcription

1 Capability requirements for a PMI research thrust fusion facility J.E. Menard, R.J. Goldston, J.M. Canik J.P. Allain, J.N. Brooks, R. Doerner, G.-Y. Fu, D.A. Gates, C.A. Gentile, J.H. Harris, A. Hassanein, N.N. Gorelenkov, R. Kaita, S.M. Kaye, M. Kotschenreuther, G.J. Kramer, H.W. Kugel, R. Maingi, S.M. Mahajan, R. Majeski, C.L. Neumeyer, R.E. Nygren, M. Ono, L.W. Owen, S. Ramakrishnan, T.D. Rognlien, D.N. Ruzic, D.D. Ryutov, S.A. Sabbagh, C. H. Skinner, V.A. Soukhanovskii, T.N. Stevenson, M.A. Ulrickson, P.M. Valanju, R.D. Woolley Columbia, LLNL, ORNL, PPPL, Purdue, SNL, UCSD, U. Ill, U. Texas ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

2 Plasma-material interface (PMI) issues and gaps drive capability requirements for the plasma integration part of a PMI thrust Access high heat flux Develop and optimize steady-state heat flux mitigation Develop and optimize transient heat flux mitigation (ELMs, disruptions) Input power / plasma surface area < 1 MW/m 2 Input power / major radius > 50 MW/m ~ Heating power / H-mode threshold power > 6, at n = n G Flexible poloidal field system for wide variation in flux expansion, ability to divert field lines to large R, liquid metals ELM/SOL regime flexibility (shape, collisionality) Stored energy / major radius ~ 5 MJ/m Non-axisymmetric coils to produce stellarator-like edge and improve MHD stability ~ High temperature ~ 1000K first wall operational Measure, control H/D/T diffusion and retention (fuel cycle, safety) Test and diagnose a range of PMI options to find optimal solution Pulse length 1000 sec; total on-time ~ 10 6 sec/yr Deuterium and trace tritium operational capability Synergy with a Fusion Materials Irradiation Facility Replaceable first wall and divertor Access for surface and plasma diagnostics, PFC services ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

3 > ~ P h /R 50 MW/m and P in /S 1 MW/m 2 are required to access DEMO-relevant heat fluxes < ~ Device R a P in P h P h /R P in /S Pulse I p Species Wall (m) (m) (MW) (MW) (MW/m) (MW/m 2 ) (sec) (MA) Temp Planned Long-Pulse Experiments EAST H (D) Low JT-60SA D Low KSTAR H (D) Low LHD ,000 H Low W7-X H (D) Low ITER DT Low Demonstration Power Plant Designs ARIES-RS Months 11.3 DT High ARIES-AT Months 12.8 DT High ARIES-ST Months 29.0 DT High ARIES-CS Months 3.2 DT High ITER-like Months 15.0 DT High EU A Months 30.0 DT High EU B Months 28.0 DT High EU C Months 20.1 DT High EU D Months 14.1 DT High SlimCS Months 16.7 DT High CREST Months 12.0 DT High P in = P aux + P α ; P h = P in -P brem -P sync For high Q systems, P h ~ (2/3) P in W/R ~ 5 MJ/m allows experiments to control ELMs and disruptions, without unacceptable PFC damage. ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

4 Design Point Scans Favor Low A 2 110keV NBI heating H 98 = 1.3 n / n gw = β N at no-wall limit (function of A) κ = κ(a) P/R and P/S goals at low size are met simultaneously at low aspect ratio ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

5 High P h / P LH is Needed to Test Highly Radiative Solution Can fusion plasmas operate with very high radiated power to reduce divertor heat flux, while maintaining good performance? Physics test requires input power exceeding H-mode threshold power by a large factor since much of the radiated power comes from within the separatrix Planned long-pulse experiments do not quite match EU-B, ARIES-RS or an ITER-like Demo EU studies EU-B: Z eff = 2.7 n/n g = 1.2 f rad = 80-90% H H = 1.2 R 0 = 8.6m I p = 28MA P = 1.33 Gw e P h /P n = n G KSTAR (29 MW) 4.1 EAST (24 MW) 5.2 JT-60SA (41 MW) 3.1 ITER (120 MW) 2.2 EU-B (653 MW) 5.8 ARIES RS (340 MW) 5.6 ITER-like Demo (400 MW) 7.3 (Y.R. Martin FEC 2004, eq. 7) ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

6 Design Point based on Minimum Electric Input Power: P/R = 50 at R 0 =1m, A=1.8 (line power) DD R0[m] 1.0 A 1.8 Ip[MA] 3.5 Bt[T] 2.0 kappa 2.7 Beta_N_total 4.5 fgw 32% fbs 62% HH P_aux[MW] 50.0 P/R [MW/m] 50 A_plasma[m^2] 43 P/S[MW/m^2] 1.15 delta 0.6 qcyl 3.47 Beta_T_total 15% Beta_P 113% ne[1/m^3] 1.10E+20 Tempavg[keV] 5.2 Flux_total[Wb] 1.9 R_inner_leg_TF [m] drfw[m] P_tf[MW] 86 P_oh[MW] 0 P_pf[MW] 38 P_aux_input [MW] 166 P_grid[MW] 300 P h /P LH = 8 (Y.R. Martin FEC 2004, eq. 7) National High-power Advanced Torus Experiment (NHTX) Systems Code Free-boundary equilibrium ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

7 Range of divertor configurations + 2D SOL/divertor calcs at P/R~50MW/m illustrate power handling challenge Poloidal flux expansion factor f exp ψ mid-plane / ψ strike-point Poloidal B-field angle of incidence into target plate α p LSN: f exp / sin(α p ) ~ 5 DN: ~ 10 DN: ~ 21 DN slot: ~ 25 ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

8 Variations in geometry strongly affect divertor parameters, but in all cases unmitigated heat flux and T e are too high Particle Flux Right Divertor Total Heat Flux Right Divertor \Gamma (1e23/m^2) : DN, fe = 21 56: DN, fe = 10 57: LSN, fe = 5 58: DN, fe = 25 q (MW/m^2) f exp / sin(α p ) LSN: ~ 5 DN: ~ 10 DN: ~ 21 DN: ~ MW/m Distance along Target (m) Electron Density Right Divertor Distance along Target (m) Electron Temperature Right Divertor ne (1E20 /m^3) Te (ev) eV P = 30MW n core = 1.5e20 R = Distance along Target (m) ReNeW Theme III PMI thrust - NHTX (Menard) Distance along Target (m) March 4,

9 Flux expansion + liquid lithium target may provide means of mitigating high steady-state heat flux At moderate evaporation rate Li vapor forms protective radiating layer With no Li evaporation, q pk = 18 MW/m 2 At 20% evaporative cooling, ~ half of the input power is radiated in the divertor Z eff q pk <6 MW/m 2 Z eff ~ 2 at plasma edge May be compatible with highperformance core plasma Also beneficial for mitigating transient heat flux from ELMs and disruptions? Lithium target ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

10 NHTX TF and VF coil set design supports very wide range of plasma, core PF coil, and divertor configurations Baseline κ=3 equilibrium and coil set κ=2 equilibrium with same coil set κ=2 equilibrium with shifted coil set κ=3 equilibrium with Super-X divertor coil set Vertical field (VF) coils Toroidal field (TF) coils Coil set compatible with 50cm neutron shield for long-duration DD hands-on access outside vessel ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

11 SOLPS modeling of NHTX with a Super-X Divertor indicate significant heat reduction is achievable Electron Density (10 20 m -3 ) Temperature (ev) Heat Flux (MW/m 2 ) Peak heat flux can be brought down to < 10 MW/m 2 Equivalent standard divertor: 34 MW/m 2 Could allow operation at low edge density (need for NBI-CD) Can be further improved with optimized target T e still high near separatrix, but much lower than the equivalent standard case (~400 ev for the blue curve) Vertical target can achieve very low T e, increase radiation ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

12 ELMs and edge turbulence, and corresponding heat flux to divertor/first-wall, can depend sensitively on plasma shape PF Coil set should support range of achievable boundary shapes NHTX: DND w/ negative squareness ζ DND w/ near zero squareness DND w/ positive squareness ζ 0.25 Example LSN shape Need to study/understand interplay between shape, q and RMP ELM control ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

13 ELM, SOL, and divertor properties depend strongly on collisionality need access to wide range of normalized plasma density NHTX accommodates high power (40-50MW) for a wide n e range: n e /n GW =0.3-1 ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

14 PMI research mission should not rely on very high plasma performance compatibility with wide range of normalized confinement, β NHTX accommodates high power (40-50MW) for a wide τ E range: H 98 = ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

15 Current drive calculations with TRANSP calculations support 0D calculations showing steady-state achievable with only NBI + BS Low-n TAE modes stable I P =3.5MA B T =2T Pedestal ν e* comparable to ITER ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

16 Double Vacuum Vessel with Removable Core Could Provide Maintainable and Flexible System Upper TF & top lid of VV removable by crane VV, divertor, inner PF coils, center stack all removable by crane TF outer & bottom legs, main VF coils fixed Cylindrical outer VV water cooled/shielded Helium cooled inner VV (first wall) at 1000k VV attachment based on access from outside Double vessel concept similar to industrial vacuum furnace technology Outside access to inner wall attachments avoids need for vessel entry after activation Prototype maintenance schemes proposed for ST-CTF and FDF? ReNeW Theme III PMI thrust - NHTX (Menard) March 4,

17 Summary A high power and highly flexible facility is needed to tame the plasma material interface for CTF & DEMO Elements of PMI research thrust: Access DEMO-relevant high heat flux for long pulses Develop steady-state heat flux mitigation Develop transient heat-flux/off-normal event mitigation Measure, control H/D/T diffusion and retention w/ hot walls Test many PMI options to find best solution(s) compatible with high plasma performance critical for attractive DEMO ReNeW Theme III PMI thrust - NHTX (Menard) March 4,