Principle Physics Developments Evaluated in the ITER Design Review Implications for NSTX

1. DRAFT Principle Physics Developments Evaluated in the ITER Design ReviewImplications for NSTX R. J. Hawrylukfor the ITER Organization, ITER Domestic Agencies, and ITER collaborators

2. Focus on Future Work • Time does not allow a “dry-run” of my ITER presentation. • Let’s talk about the implications for NSTX • Research opportunities identified in a box • Some but not all are a good match for NSTX • No attempt to be exhaustive here or address all of the ITER high priority topics • Comments are not meant to be prescriptive but give my impressions.

3. Objective: Update Physics Requirements • Focus of the talk is on the impact of recent physics results affecting the ITER design with emphasis on near-term procurement arrangements • Confinement (sensitivity studies) • Plasma shaping and vertical stability • TF ripple • First wall design • ELM control: pellet pacing and RMP coils • RWM control • Disruption and disruption mitigation

4. Reliable Operation Is Needed to Meet Mission Goals • Performance in ELMy H-mode is defined by: - confinement ( assumed H-mode scaling) - L to H and H to L power threshold - density - auxiliary heating power • Baseline case, 15MA, 5.3T Q ~10 at ne/neG = 0.85, HIPB98(y,2) =1 - 13.5MA, 4.77T Q ~6 - 17MA, 5.3T Q ~20 - 10% reduction in k at constant q95, Q~6 • Reinforced the importance of reliably operating ITER at full Bt, Ip, and k. - Research on advanced operating modes and - Techniques to decrease power threshold. 15 MA, 5.3T 13.5MA, 4.77 T J. Johner

5. ITER Demonstration Discharges Used to Simulate Startup and Evolution • Adopted large aperture startup, early heating, and divertor attachment to decrease li in the startup phase. • li decreases to as low as 0.6 in flattop phase. • li increases during (deliberate) H to L transition and current shutdown. li HL HL A.C. C. Sips et al. IT-2-2

6. Performance of ITER PF System Was Evaluated • Analysis showed that operating space was not adequate for H-modes with large pedestals. C. E. Kessel et al. IT/2-3

7. Design Changes Enable Low Inductance (H-mode) operation • Increase the current and field capability of the PF conductor; • Increase the number of turns in PF2 and PF6; • Increase the limit on the central solenoid vertical separation forces (from 75 MN to 120 MN); • Relocate PF6 toward the plasma by 9 cm and radially by 7 cm; • Sub-cool PF6 to about 3.8 K; and • Modify the divertor slots and dome geometry • Current analysis is focusing on analyzing the effect of plasma disturbances on the operating range and a • Detailed assessment of rampdown phase of the discharge including the H to L transition. C. E. Kessel et al. IT/2-3.

8. Vertical Position Control Must Be Robust and Reliable in ITER • Loss of vertical plasma position control in ITER will cause large thermal loads on PFCs • VDE generate the highest electromagnetic loads. • Experiments on C-Mod, DIII-D, JET, NSTX, and TCV have provided a criteria for evaluating the vertical stability control: • Dz/a >0.05 for reliable vertical stability • Dz/a >0.1 for robust vertical stability • Original system capable of Dz/a ~0.02. • Evaluating design of internal coils for vertical stability, ELM and RWM control. • Capable of Dz/a >0.05 Upper VS coil Upper ELM coil mid-plane ELM coil Lower ELM coil Lower VS coil Noise in diagnostics A. Portone et al. IT/2-4Ra; D. Humphreys et al. IT/2-4Rb.

9. Reduction in Toroidal Field Ripple Increases Energy Confinement Time JT-60U • Degradation in tE, pedestal height, and plasma rotation observed with increasing TF ripple. • Reduce ripple to “as low as reasonably achievable” was approved. • Underlying physics of how ripple affects tEis under study. • What are the implications for TBM requirements? • Is ripple contributing to the loss of fast particles in NSTX? Urano, et al. (2006) JET F. Saibene et al EX/2-1

10. Heat Load on First Wall Impacted the Design • Parallel heat fluxes in scrapeoff layer dominated by intermittent events, which are characterized by radial velocity. • Parallel heat fluxes beyond the second separatrix are estimated. • To avoid damage to the edges of the blanket shield module • First wall shape has been modified. • Modules, in toroidal location of port plugs, are recessed. • Heat loads are not well known to divertor or first wall A. Loarte et al., IT/P6-13. C. G. Lowry et al. IT/1-4.

11. Unmitigated ELMs Will Limit Divertor Lifetime J. Linke, et al. (2007) • Recent results show unmitigated ELMS correspond to ~10MJ/m2. • Evaluated two approaches: pellet pacing and suppression by resonant magnetic perturbations (RMP). P. R. Thomas et al. IT/1-5.

12. Pellet Injection Used to Trigger More Frequent Smaller ELMs • Asdex Upgrade has reduced the ELM size by a factor of ~1.6 by pellet pacing with a small decrease in tE. - Attributed to increased convective loss. • Need to reduce the ELM size by ~20 in ITER with <10% reduction in tE. • Increased gas load requirements to accommodate pellet pacing. • Need further research on: - Depth of pellet penetration required to trigger an ELM. - Development of higher speed pellet injector. - Not planned on NSTX P. T. Lang et al. (2002)

13. Resonant Magnetic Perturbations Suppressed ELMs in DIII-D • Suppressed ELMs with n=3 RMPs, with small aperture off-mid-plane coils • Obtained HIPB98(y,2)=1 • 3.2<q95<3.8 • Density decreased • Incorporated into the design of the in-vessel coils based on DIII-D results and theoretical considerations. • Understanding of the underlying physics is still emerging. • Criteria for field line alignment and mode spectrum • Role of edge pumping • Effectiveness of core pellet fueling M.E. Fenstermacher et al. (2008)

14. Control of Resistive Wall Modes (RWM) Enables Steady-state Operating Scenarios • “Steady-state” operation in ITER entails bN>3, which can result in a RWM. • Even if rotation can stabilize the RWM, can be excited by finite amplitude error fields or other MHD activity. • Active feedback control on DIII-D and NSTX have shown that it is possible to stabilize RWM even at low rotation. • In-vessel coils are predicted to stabilize RWM to bN>3.8. • Coil current requirements are modest. • Further analysis and benchmarking of codes is in progress. VALEN Code J. Bialek

15. Revised Disruption Loads Accommodated in the Vacuum Vessel Support Structure • Largest disruption loads are due to VDE • Peak downward vertical forces revised from 75 to 108 MN • JET observes large toroidal asymmetry of plasma current and Zp and resulting large sideways force. • Peak horizontal force revised from 25 to 50 MN • Understanding the underlying physics and improving the extrapolation to ITER remain active areas of research. JET V. Riccardo et al. 2000

16. Massive Gas Injection Has Been Used for Disruption Mitigation on Present Experiments • Resulted in short current decay time and radiative loss of plasma and poloidal magnetic energy. • Detection of VDEs should be reliable, and mitigation possible, due to long ITER timescales. • Necessary part of PFC/FW protection. • The current and major radius of ITER is a substantial extrapolation from existing machines. • Avalanche generation of runaway electrons is predicted if density is less than Connor-Hastie-Rosenbluth density. • Collisional damping requires a gas influx of 500 kPa•m3 assuming a 20% fueling efficiency • Large impact on vacuum and tritium systems • Workshop was held in July and identified research areas including: • Are the runaways well confined, requiring collisional damping? • How should the gas, liquid or pellets be injected? D. G. Whyte et al. IT/P6-18

17. IO Working with the Scientific Community Has Advanced the ITER Design • Key issues affecting the procurement agreements were addressed: • Vacuum vessel and blanket shield module design • In-vessel coils • Poloidal field coil systems • Identified important scientific and technical questions, which require further experimental and theoretical work to support the design and research operations. • Disruption and runaway electron mitigation. • ELM control and suppression • Heat fluxes to plasma facing components. • Continued close interaction between the IO and the scientific and technical community is critical to ensure that optimal use is made of ITER.

18. What Has Not Been Resolved? • Choice of first wall material • Solomon-like decision: use everything at startup • Good results with W from Asdex Upgrade except with ICRF • How to get the tritium out? • Replacing the divertor with tungsten delays the DT schedule but a viable plan for tritium removal does not exist. • Measurement and removal of dust • Steady-state and hybrid operating modes • Can we count on mystery mechanisms to maintain high q(0)? • Broader issue of resonant and non-resonant perturbations. • Can we make NNBI work?