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22nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 1
Principle Physics Developments Evaluated in the ITER Design Review
Implications for NSTX
R. J. Hawrylukfor the ITER Organization, ITER Domestic
Agencies, and ITER collaborators
DRAFT
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22nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 2
• 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.
Focus on Future Work
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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
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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 at constant q95, Q~6
• Reinforced the importance of reliably operating ITER at full Bt, Ip, and
- Research on advanced operating modes and- Techniques to decrease power threshold.
€
τE ,thIPB 98(y,2) = 0.0562HIPB 98(y,2)Ip
0.93BT0.15n e
0.41P−0.69R1.97M 0.19κ a0.78ε 0.58
J. J. JohnerJohner
15 MA, 5.3T
13.5MA, 4.77 T
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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.
0.6
0.8
1
1.2
0 1 2
0.6
0.8
1
1.2
0 1 2 3 4
AUG DIII-D
Time (s) Time (s)
0.6
0.8
1
1.2
0 5 10 15
JET
Time (s)
FTrise rise rise
FT
FT
FT
FT
(a) (b) (c)
0.6
0.8
1
1.2
0 1 2
0.6
0.8
1
1.2
0 1 2
0.6
0.8
1
1.2
0 1 2
0.6
0.8
1
1.2
0 1 2 3 4
0.6
0.8
1
1.2
0 1 2 3 4
0.6
0.8
1
1.2
0 1 2 3 4
AUG DIII-D
Time (s) Time (s)
0.6
0.8
1
1.2
0 5 10 15
JET
Time (s)
0.6
0.8
1
1.2
0 5 10 15
JET
Time (s)
FTrise rise rise
FT
FT
FT
FT
(a) (b) (c)
ll ii
A.C. C. Sips A.C. C. Sips et al. et al. IT-2-2IT-2-2
HLHL
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Performance of ITER PF System Was Evaluated
• Analysis showed that operating space was not adequate for H-modes with large pedestals.
C. E. Kessel C. E. Kessel et al. et al. IT/2-3IT/2-3
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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.
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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:
z/a >0.05 for reliable vertical stability z/a >0.1 for robust vertical stability
• Original system capable of z/a ~0.02.• Evaluating design of internal coils for vertical
stability, ELM and RWM control.– Capable of z/a >0.05
Lower ELM coil
Upper ELM coil
Upper VS coil
mid-planeELM coil
Lower VS coil
A. PortoneA. Portone et al. et al. IT/2-4Ra; D. Humphreys IT/2-4Ra; D. Humphreys et al. et al. IT/2-4Rb.IT/2-4Rb.
Noise in diagnostics
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Reduction in Toroidal Field Ripple Increases Energy Confinement Time
• Degradation in τE, 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 τEis under study.
– What are the implications for TBM requirements? – Is ripple contributing to the loss of fast particles in NSTX?
Urano, et al. (2006)
JT-60U
JET
F. Saibene et al EX/2-1
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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.
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•• Recent results show unmitigated ELMS correspond to ~10MJ/mRecent results show unmitigated ELMS correspond to ~10MJ/m22..•• Evaluated two approaches: pellet pacing and suppression by Evaluated two approaches: pellet pacing and suppression by
resonant magnetic perturbations (RMP).resonant magnetic perturbations (RMP).
Unmitigated ELMs Will Limit Divertor LifetimeUnmitigated ELMs Will Limit Divertor Lifetime
P. R. Thomas et al. IT/1-5.
J. Linke, et al. (2007)
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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 τE.- Attributed to increased convective loss.
• Need to reduce the ELM size by ~20 in ITER with <10% reduction in τE.
• 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)
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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)
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Control of Resistive Wall Modes (RWM) Enables Steady-state Operating Scenarios
• “Steady-state” operation in ITER entails N>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 N>3.8.
– Coil current requirements are modest.• Further analysis and benchmarking of
codes is in progress.J. Bialek
VALEN Code
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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
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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
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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.
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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?