Alcator C Mod. Integrated Scenarios ITER H-mode Baseline. Presented by: Stephen M. Wolfe

Transcription

1 Integrated Scenarios ITER H-mode Baseline Presented by: Stephen M Wolfe C-Mod PAC Meeting MIT Plasma Science & Fusion Center Cambridge, MA Feb 5, 2009

2 Definitions and Scope By Integrated Scenarios, we mean research aimed at reaching attractive operating points, generally cutting across multiple science topics and often involving interaction and compatibility issues between different plasma processes or regions The ITER Q = 10 DT H-mode baseline scenario features Positive shear, q 0 < 1 q 95 3, β N = 18, H H 1, f NI 025 Edge transport barrier Research effort also supports aspects of the ITER Pre-nuclear Commissioning Phase Operation in H and/or He majority Reduced parameters ( 1 2 B T, I p ) Alternate RF Scenarios Issues of H-mode access, character C-Mod PAC Meeting Feb 5, 2009 smw 1

3 Research Topics Compared to the Topical Science areas, the H-mode Integrated Scenarios Research Program is more focussed on support of ITER construction and operations planning Respond to requirements identified by IO Participate in identification, definition, and execution of Joint Experiments and High Priority research activities of ITPA Proactively identify and address issues relevant to ITER planning and operation Address C-Mod issues of integration and scenario development impacting our ability to carry out ITER-related research C-Mod PAC Meeting Feb 5, 2009 smw 2

4 Relevance of C-Mod research for ITER C-Mod physics regime, machine capabilities and control tools are uniquely ITER-relevant in many respects: Edge and Divertor: All metal walls High divertor heat fluxes 05GW/m 2 High SOL neutral opacity (similar fueling to ITER) High Lyman opacity, radiation trapping (closest to ITER) Core Transport: Equilibrated ions and electrons No core fueling or momentum sources (very low on ITER) Macro-stability Can access ITER β range at same B T and absolute pressures Wave Physics: Similar tools (ICRF and LHCD) to ITER Same B,n ω p, ω c, similar ω (key for waves, LH feasibility) Pulse length: τ pulse >> τ cr (exceeds ITER) Non-inductive CD capability Combination of these features is unique and is key to Integrated Scenarios C-Mod PAC Meeting Feb 5, 2009 smw 3

5 Addressing ITER H-mode Scenario issues requires integration Issues, Challenges are similar for C-Mod and ITER Divertor and wall materials and conditioning High heat flux requirement Compatibility with low core radiation, Z eff Hydrogen retention and recovery ICRF Heating Challenging RF power density Compatibility with H-mode edge, high n e operation, ELMs Compatibility with wall conditioning requirements Control of pedestal parameters, Edge relaxation Variation of edge parameters (ν,β) Particle and impurity control ELM control C-Mod PAC Meeting Feb 5, 2009 smw 4

6 C-Mod characteristics relevant to ITER operational studies C-Mod & ITER PF sets have similar arrangement and (multi-parameter) functionality Large vessel and structure currents, especially during startup, ramp-down MIMO shape control with few actuators, minimal null space, operation near current, voltage, stress limits RF-based actuators for heating, non-inductive current drive Negligible central particle, momentum sources C-Mod PAC Meeting Feb 5, 2009 smw 5

7 Modeling and Simulation capability is key for Integrated Scenario Development Discharge operational design facilitated by Alcasim code (M Ferrara) MATLAB/Simulink application developed by MIT grad student Coupled to DPCS plasma control system software and operator interface Detailed model of C-Mod sensors, power system actuators Simplified plasma model Parameter evolution based on experimental data or simulations Use/Validate codes such as TSC to advance tokamak discharge subject to transport estimates from TGLF or GLF23 (PPPL Collaboration) Run TRANSP on time advanced discharge from TSC to compute source terms due to ICRF (TORIC-FPP), LHCD (LSC code being upgraded to modules based on CQL3D/GenRAY) Perform time advance in TSC using new source terms from TRANSP calculation Applications to experiment design and interpretation Use sophisticated simulation models in standalone mode developed through SciDAC Projects and Base Program (see talk by Bonoli): NIMROD and M3D - Extended MHD (AORSA, TORIC) - (CQL3D, ORBIT RF) for ICRH CQL3D - GENRAY for LHCD XGC, GS2 for transport C-Mod PAC Meeting Feb 5, 2009 smw 6

8 H-mode Scenarios experiments in FY2008 Approximately 75 run days in 2008 (12% of total) High performance in ITER-like equilibria with q 95 4 (I p > 12 MA) ELM studies C-Mod/NSTX/MAST small ELM regime comparison (PEP-16) Scaling of L-H threshold power at low density (CDB-11) Access to reduced collisionality H-mode Development of ITER relevant current rise and ramp-down phase (Sips) (SSO-5) C-Mod PAC Meeting Feb 5, 2009 smw 7

9 Validation of ITER reference startup and rampdown scenarios (Sips, IPP-Garching/ITPA SSO-5) ITER startup scenario requires low 07 < l i < 1 for vertical stability Proposed ramp to 15 MA flattop in 80 sec scales to C-Mod 135MA in 035 to 06 sec (depending on T e ) Full bore plasma with x-point formation at 1 4 maximum current needed to avoid overheating ITER limiters Demonstration of satisfactory ITER rampdown, termination scenario Control of divertor shape, l i excursion, flux consumption Maintenance of robust vertical stability Safe termination of H-mode, burn phase Control over particle inventory 14 Ip (MA) li (3) Te0 (kev) Dsep (cm) Prf (MW) 2 ITER-like Startup Experiments t (sec) Impacts PF coil, Power supplies, Decision on Internal Coils C-Mod PAC Meeting Feb 5, 2009 smw 8

10 Overview of C-Mod results l i 05s (end of current rise) TSC/TRANSP resistive rise flat top inductive Ohmic ICRH V-s required Ohmic 2MW ICRH T e0 [kev], av 03-05s Time [s] During divertor phase (03-05s) higher T e lower l i (3) TCS/TRANSP: Heating mainly saves resistive flux during current rise The inductive flux is stored and used by the plasma (same for ITER) George Sips APS Conference, Dallas, November 17-21,

11 Current ramp down experiments Ramp down in divertor shape with kappa reduction (18 14) 132MA 098MA 061MA 026MA 2MA/s 114 s 134 s 154 s 172 s time (s) George Sips APS Conference, Dallas, November 17-21,

12 Current ramp down experiments k k κ 14 κ 14 I p [MA] I OH [ka] 10% κ~18 l i (3) ( s ) Time Elongation reduced from 18 to 14 to keep plasma vertically stable Current ramp down keeping κ~18 has VDE For ohmic or L-mode discharges, fast decay (di p /dt ~ 4MA/s) or standard decay (di p /dt ~ 2MA/s) l i (3) > 15 1MA/s ramp down keeps l i (3) < 12 However, this ramp down requires ~10% increase in central OH current George Sips APS Conference, Dallas, November 17-21,

13 Emphasis of future program on issues related to ITER construction and operation Demonstration of Scenario 2 - like equilibrium, p, T e /T i, power density (IOS-11) Control of operating point Benchmark transport, stability modeling Evaluate impurity, particle control Evaluate disruptivity, stability issues Exploration of variations about reference scenario Continue Validation of robust ramp-down, shut-down sequence (IOS-22) Compatibility of core and boundary Interaction with plasma-facing materials, particle control, impurity seeding IOS-12) including heat-flux and Demonstration of reliable, benign fault-handling and mitigation techniques Address issues associated with initial ITER operating phase C-Mod PAC Meeting Feb 5, 2009 smw 12

14 ITER Pre-Nuclear Phase (first 2-3 years) No deuterium operation majority H (or He), minority He 3 Heating, CD sources under commissioning P tot < 73MW (NBI, EC, IC) 40 < f ICRF < 55MHz, P ICRF 20MW B T between half and full field CFC divertor (decision on transition to W tbd) If this isn t possible, higher current operation (consistent with the requirements for ECR ICRH) may be required For the purposes of this plan, it is assumed that plasma condition be sustained for up to s, allowing commissioning of all available heating systems same pulse The suggested allowance is between 72 and 108 successful operational days longer time would allow time to commission the ICRH to high C power Modfor both He3 mino full field and H minority (in He4) at reduced field At present the commissioning of th beam line with plasma to full power (short pulse) would continue into the next opera phases Fig 321 Available operation space for various heating systems on ITER in terms of plasma c toroidal field, and plasma density The figure shows the plasma current (left axis) and density axis) against toroidal field The three diagonal lines correspond to 80% of the Greenwald lim q 95 = 3, 4 and 5 The green-shaded areas correspond to fundamental and 2 nd harmonic He 3 minor Issues Cyclotron Resonance Heating (ICRH) for the available frequency range The higher field ban corresponds to 2 nd harmonic T The blue-shaded area corresponds to H minority ICRH (see text Access to H-mode yellow bands correspond to the accessible domain for the main Electron Cyclotron Resonance H (ECRH) system at the fundamental and 2 nd harmonic The bands are vertically displaced to sho overlap between ECRH and ICRH The horizontal red line at m -3 and the dashed red Alternate RF scenarios (eg 2ω m -3 correspond to the estimated Neutral Beam Injection (NBI) shinethrough limits for 08 He 3 at 27 T) for the cases with and without shinethrough armour Note that simultaneous NBI and ECH commissioning are not compatible with Plasma control at q 95 5, L-mode commissioning (l i > 1 over?) the entire frequency range of ICRH operation Of particular releva the low frequency range of ICRH operation (ie, lower field in Fig 321) in which operation is not possible and NBI operation will require operations at high densities relat C-Mod PAC Meeting Feb 5, 2009 smw 13 the Greenwald density In these cases, specific time will be devoted to ICRH commissi While present-day experiments suggest that ICRH heating 2 He3 is possible, it is antic that parasitic losses due to the absorption at other prominent harmonics (in particular,

15 Proposed Near-term C-Mod Activities in support of pre-nuclear phase Investigate L-H threshold low-density limit (TC-3) initial experiments completed species dependence of transition and confinement (TC-4) He (and H?) majority Characterize He majority H-modes Limited experience in C-Mod indicates differences in pedestal relaxation, divertor characteristics, impurity transport Extend systematic studies to ITER-relevant regimes, benchmark simulation models Assess relevant ICRF scenarios He 3 minority (first and second harmonic) Consider H majority scenarios Assess ITER ramp-up, flattop, ramp-down scenarios in L-mode, q 95 5 C-Mod PAC Meeting Feb 5, 2009 smw 14

16 Extend high performance plasma studies at ITER field, shape, I/aB, β, Integrated tests of confinement, heating, power handling Operate at ITER field (53T), maximum power (P ICRF > 5MW ) f 80MHz, D(H) heating high single pass absorption High elongation (κ = 18) ITER shape fills C-Mod vessel Increase current to I p = 16MA with q 95 > 3 ITER Pressure Use cryo-pump for improved density control, lower collisionality Expanded parameter space for databases and extrapolation to ITER Assume W α I p * Demonstration of operation at ITER absolute pressure C-Mod PAC Meeting Feb 5, 2009 smw 15

17 Range of collisionalities required for ITER H-mode validation experiments Because ρ is not matched, no single measure of collisionality is adequate to characterize different physical processes Some transport effects may be characterized by ν neo = ɛ 3/2 ν ii qr/v thi Others, along with NTM effects, depend on ν/ω, which is larger by a factor ρ Electron-ion equilibration depends on ν e/i τ E, depends on ρ scaling of τ E SOL-Divertor interaction may depend on n/n G Ratio to ITER ITER Demonstration Discharges υ * neo υ/ω * υ e/i τ E gb β N =18 B=53 T P CMod (MW) [H 89 =2] 0 n CMod /n G n (10 20 m -3 ) It is not in general possible to model all the relevant physical processes in the same demonstration discharge Near-term ITPA Joint Experiment IOS-11 aims for n/n G 085 C-Mod PAC Meeting Feb 5, 2009 smw 16 ρ *

18 centrations in the core This will reduce the detrimental effects on the core such as fuel dilution and cooling, and will enhance the positive effects of the dissipative divertor Two measures of divertor retention are impurity compression and Operational impurity control enrichment of divertor power balance 40,41 Impurity compression is defined as C z n div 0,z /n core z, where n div 0,z is the divertor impurity neutral density and n using radiation seeding core z is the core impurity ion density Maximiz- (IOS-12) ing compression can result in more power exhausted through radiation in the divertor if the temperature is optimal Also, with n div 0,z constant, maximizing compression leads to less contamination of the core plasma Impurity enrichment is Goal is to reduce divertor heat load by increasing defined as z C z /C D, where C D is the deuterium compression Impurity radiative loss without compromising particle exhaust enhanced pedestal, by enriching core the impurity level with respect to the fuel gas in the divertor confinement (ITER Measurements scenario) of the impurity compression and enrichment have been made for a variety of discharge conditions C-Mod experiments feature ITER-like divertor 42 including H-modes with a detached outer divertor Argon is normally puffed in trace amounts for central ion temperature power density, and geometry, rotation measurements neutral In anopacity, attached EDA H-mode Extensive pedestal and divertor diagnostics plasma, C Ar is 50 and Ar is 5 Parallel deuterium ion flows toward the divertor in the SOL are measured 43 and appear to play a role in keeping the compression high After the outer divertor becomes detached, C Ar and Ar fall a fac- Feedack controlled tor of two Theimpurity change in compression puffing and enrichment using is due mostly to a change the screening of the impurity from the recycling and non-recycling impurities core plasma 32 The amount of neutral impurity in the divertor falls only slightly after the divertor detaches This decrease Cryopump provides in compression sink and enrichment for better should be regulation, considered as a disadvantage of detached divertor operation 4 control of impurity level C Radiation effects Begin (resume) experiments (Mo divertor) Attempts have been made to detach H-mode plasmas using neon and argon as the puffed impurity gas 8 Unlike nitrogen, these are fully recycling gases In contrast to the nitrogen injection experiments, these attempts were not successful in producing a detached divertor Table I lists the Continue with high heat-flux, symmetric W divertor ( ) cient than a decrease in the power flowing into the SOL i reducing B Plasma thecharacteristics local divertor temperature and thus inducin detachment This is a possible explanation of why nitroge A comparison of EDA H-mode plasmas with and with gas puffing in C-Mod, which out a detached outer divertor isc detailed increases P Mod div in Figs 3 rad and de and 4 Th creases P EDA H-mode SOL, is more effective in inducing detachment tha plasma without impurity puffing is denote neon and argon, which only decrease P by the dashed lines and the dissipative SOL divertor H-mod Core plasma performance is affected by neon and argo plasma with impurity puffing by the solid lines in bot puffing Enhancement factors for H-modes with neon an argon puffing are degraded In fact, the H ITER89P for thes attached cases is below that of detached H-modes with nitro gen puffing 8 Experiments with neon and krypton puffin have been done to determine the effect of these gases on th edge temperature pedestal 44 The temperature pedesta clearly decreased in height, as well as width, as the cor radiation increased and H ITER89P decreased correspondingly Measured radiation emissivity profiles due to these gases ar shown in Fig 8 a The profile for nitrogen is also shown I is seen that neon and krypton radiate inboard of the pedesta region shaded and that nitrogen radiates farther out Emis sivity profiles obtained from modeling with the MIST impu rity transport code 45 support these measurements Fig 8 b FIG 4 Typical divertor plasma conditions are shown for the discharges Fig 3 Downloaded 28 Apr 2008 to Redistribution subj FIG 8 Radiation emissivity profiles are measured a and calculated wit the MIST impurity transport code b for impurities puffed into an ED H-mode plasma The pedestal region is shown with the shading change in radiation emissivity measured in the edge of the main plasma and in the divertor plasma that is induced by nitrogen, neon, and argon puffing It is seen that all gases increase the radiation from the edge plasma which in turn decreases the inferred power flowing into the SOL and divertor Little or no increases in divertor radiation were ob- C-Mod PAC Meeting Feb served 5, when 2009 using neon and argon smw 17 Downloaded 28 Apr 2008 to Redistribution subject to AIP license or copyright; see

19 Development and testing of plasma control algorithms Testing of plasma control strategies in relevant scenarios Effects of sensor noise, transient events Effects of currents in passive structures on control and reconstructions Benchmarking of simulation codes Development and testing of machine protection algorithms Identification of and response to sensor, actuator faults Response to proximity to actuator limits adaptive transfer to safe shutdown sequence or real-time pulse rescheduling Real-time identification of proximity to plasma instability boundaries Implementation of disruption mitigation algorithms in routine tokamak operation for evaluation of robustness and reliability C-Mod PAC Meeting Feb 5, 2009 smw 18

20 Burn Control simulation experiments with ICRH 1 Study evolution and stationary states of plasma with power dependent on plasma parameters Use ICRF minority heating to mimic centrally peaked fast ion heating Control (part of) P ICRF n 2 f(t ) or R DD to simulate burn Apply feedback to try to maintain constant burn power against perturbations such as ELMs, sawteeth, MHD, density excursions, etc 1 suggested by P Politzer (GA) C-Mod PAC Meeting Feb 5, 2009 smw 19

21 H-mode Integrated Scenario research supports ITER/ITPA Joint Experiments Description Joint Experiment Notes on C-Mod Contributions L-H threshold power at low density TC-3 (CDB- Initial C-Mod experiment completed 11) (2007) H-mode transition and confinement TC-4 (CDB- for ITER pre-nuclear phase dependence on ionic species 12) begin 2009 Scaling of spontaneous rotation with no TC-9 (TP61) (2008) external momentum input C-Mod, NSTX, MAST Small ELM regime PEP-16 (2008) comparison Non-resonant magnetic braking MDC Vertical Stability Physics and performance limits in Tokamaks with highly elongated MDC-16 Report submitted, additional work pending plasmas Simulation and validation of ITER Startup to achieve advanced scenarios SSO-5 Addressing both AT and H-mode baseline scenarios (2008?) ITER demo at q 95 = 3,β N = 18,n e 085n G IOS-11 New 2009 Study seeding effects on ITER demo IOS-12 New Begin 2009 discharges Ramp-down from q 95 = 3 IOS-22 Continue experiments begun under SSO-5 C-Mod PAC Meeting Feb 5, 2009 smw 20

22 H-mode Scenarios Priorities ITER demo (IOS-11) and related scenarios 2009: access β N 18, q 95 3, at moderate n/n G (EDA) 2010: Extend to higher and lower collisionality as feasible 2011: Continue exploration of Scenario 2-like operating space Impurity seeding, divertor power balance, core compatibility 2009: Initial experiments on IOS-12, maybe open-loop?, species optimization 2010: Demonstration with feedback, optimize 2011: Continue with high heat-flux symmetric divertor ITER rampdown studies IOS-22 ( ) ITER pre-nuclear phase issues 2009: Evaluate He 4 H-mode threshold, characteristics : Evaluate H majority scenarios (beginning or end of campaign) : (as needed) address ramp-up, ramp-down, flattop control issues for reduced parameter conditions (L-mode, q 95 5) Development and testing of plasma control algorithms 2009: routine use of mitigation algorithms? : advanced control, fault sensing, adaptive algorithms Burn control studies (2010? or 2011) C-Mod PAC Meeting Feb 5, 2009 smw 21

23 C-Mod Research on Integrated Scenarios for ITER H-mode Baseline H-mode baseline research program addresses cross-cutting physics and technology issues Exploits ITER-relevant C-Mod parameters and tools Strongly coupled to ITPA tasks, Joint Experiments Supports planning for ITER pre-nuclear operating phase Many additional ITER-related experiments in Topical Science Groups Integrated Research in support of ITER Advanced Scenarios described in Next Presentation C-Mod PAC Meeting Feb 5, 2009 smw 22