AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc NSO Collaboration Thoughts on Fusion Development Dale Meade Princeton Plasma Physics Laboratory FPA Annual Meeting December 13, 2004

We will never get there if we don’t take the next step. Thoughts on the Magnetic Fusion Development Path A plan was put forward by FESAC (Freidberg Panel) to establish a national technical consensus on a Next Step for Magnetic Fusion - A Burning Plasma Program is needed. A solid technical basis exists to design and construct a tokamak BPX. A BP strategy and Development plan were developed with over whelming support of fusion program leaders, unanimous endorsement by FESAC and put forward to DOE. The technical basis for the strategy and plan is even stronger today. We should follow this plan - do not go back to Square One!

FESAC/DOE Plan for Developing Fusion Energy If no ITER, then go with FIRE which leads into an advanced DEMO.

High Power Density P f /V~ 6 MWm -3 ~10 atm  n ≈ 4 MWm -2 High Gain Q ~ n  E T ~ 6x10 21 m -3 skeV P  /P heat = f  ≈ 90% Low rotation Steady-State ~ 90% Bootstrap ARIES Economic Studies have Defined the Plasma Requirements for an Attractive Fusion Power Plant Plasma Exhaust P heat /R x ~ 100MW/m Helium Pumping Tritium Retention Plasma Control Fueling Current Drive RWM Stabilization Significant advances are needed in each area. A Burning Plasma experiment is to make progress in these areas.

Advanced Toroidal Physics (100% Non-inductively Driven AT-Mode) Q~ 5 as target, higher Q not precluded f bs = I bs /I p ~ 80% as target, ARIES-RS/AT≈90%  N ~ 4.0, n = 1 wall stabilized, RWM feedback Quasi-Stationary Burn Duration (use plasma time scales) Pressure profile evolution and burn control>  E Alpha ash accumulation/pumping>  He Plasma current profile evolution~ 2 to 5  skin Divertor pumping and heat removal>  divertor First wall heat removal> 1  first-wall FIRE Physics Objectives Burning Plasma Physics (Conventional Inductively Driven H-Mode) Q~10 as target, higher Q not precluded f  = P  /P heat ~ 66% as target, up to Q = 25 TAE/EPMstable at nominal point, access to unstable

Fusion Ignition Research Experiment (FIRE) R = 2.14 m, a = m B = 10 T, (~ 6.5 T, AT) I p = 7.7 MA, (~ 5 MA, AT) P ICRF = 20 MW P LHCD ≤ 30 MW (Upgrade) P fusion ~ 150 MW Q ≈ 10, (5 - 10, AT) Burn time ≈ 20s (2  CR - Hmode) ≈ 40s (< 5  CR - AT) Tokamak Cost = $350M (FY02) Total Project Cost = $1.2B (FY02) 1,400 tonne LN cooled coils Mission: to attain, explore, understand and optimize magnetically-confined fusion-dominated plasmas

FIRE is Based on ARIES-RS Vision 40% scale model of ARIES-RS plasma ARIES-like all metal PFCs Actively cooled W divertor Be tile FW, cooled between shots Close fitting conducting structure ARIES-level toroidal field LN cooled BeCu/OFHC TF ARIES-like current drive technology FWCD and LHCD (no NBI/ECCD) No momentum input Site needs comparable to previous DT tokamaks (TFTR/JET). T required/pulse ~ TFTR ≤ 0.3g-T

Extended H-Mode and AT operating ranges Benefits of FIRE high triangularity, DN and moderate n/n G Extended H-Mode Performance based ITPA scaling with reduced  degradation, and ITPA Two Term (pedestal and core) scaling (Q > 20). Hybrid modes (AUG, DIII-D,JET) are excellent match to FIRE n/n G, and projects Q > 20. Slightly peaked density profiles (n(0)/ = 1.25)enhance performance. Elms mitigated by high triangularity, disruptions in new ITPA physics basis will be tempered somewhat. Significant Progress on Existing Tokamaks Improves FIRE (and ITER) Design Basis since FESAC and NRC Reviews

New ITPA Scaling Opens High-Q H-Mode Regime for FIRE Systematic scans of  E vs  on DIII-D and JET show little degradation with  in contrast to the ITER 98(y, 2) scaling which has  E ~  A new confinement scaling relation developed by ITPA has reduced adverse scaling with  see eq. 10 in IAEA-CN-116/IT/P3-32. Cordey et al. A route to ignition is now available if high  N regime can be stabilized.

FIRE AT Mode Operating Range Greatly Expanded Q = 5 Nominal operating point Q = 5 P f = 150 MW, P f /V p = 5.5 MWm -3 (ARIES) ≈ steady-state 4 to 5  CR Physics basis improving (ITPA) required confinement H factor and  N attained transiently C-Mod LHCD experiments will be very important First Wall is the main limit Improve cooling revisit FW design Opportunity for additional improvement.

“Steady-State” High-  Advanced Tokamak Discharge on FIRE P f /V = 5.5 MWm -3  n ≈ 2 MWm -2 B = 6.5T  N = 4.1 f bs = 77% 100% non-inductive Q ≈ 5 H98 = 1.7 n/n GW = 0.85 Flat top Duration = 48  E = 10  He = 4  cr FT/P7-23

TAE Modes in FIRE AT Regime Nova-K: N. Gorelenkov   (0) = 0.62%, R   = 3.34 %, T(0) = 14 keV n = unstable radially localized just inside q = 0.8 modes are weakly unstable due to low alpha-particle beta

FIRE has Passed DOE Physics Validation Review The DOE FIRE Physics Validation Review (PVR) was held March in Germantown. The Committee included: S. Prager, (Chair) Univ of Wisc, Earl Marmar, MIT, N. Sauthoff PPPL, F. Najmabadi, UCSD, Jerry Navratil, Columbia (unable to attend), John Menard PPPL, R. Boivin GA, P. Mioduszewski ORNL, Michael Bell, PPPL, S. Parker Univ of Co, C. Petty GA, P. Bonoli MIT, B. Breizman Texas, PVR Committee Consensus Report: The FIRE team is on track for completing the pre-conceptual design within FY 04. FIRE would then be ready to launch the conceptual design. The product of the FIRE work, and their contributions to and leadership within the overall burning plasma effort, is stellar. Is the proposed physical device sufficiently capable and flexible to answer the critical burning plasma science issues proposed above? The 2002 Snowmass study also provided a strong affirmative answer to this question. Since the Snowmass meeting the evolution of the FIRE design has only strengthened ability of FIRE to contribute to burning plasma science.

Steps to a Magnetic Fusion Power Plant FIREARIES-RS ITER FIREARIES-RS Fusion Gain10(H), 5(AT) 25 (AT) Fusion Power (MW) Power Density(MWm -3 ) Wall Loading  n (MWm -2 ) Pulse Duration (s) (  CR, %) , 86 - >99.9% , 86 - >99% 20,000,000 steady Mass of Fusion Core (tonnes)19,0001,40013,000

FIRE FIRE Physics Validation Review successfully passed. March 30-31, 2004 FIRE Pre-Conceptual Activities are completed. September 30, 2004 Ready to begin Conceptual Design Activities. Now “Hold our FIRE” as per Fusion Energy Sciences Advisory Committee recommendation ITER AT Extend performance of ITER using Advanced Tokamak operation Fully exploit the capability of ITER (increase power to ~1GW at steady-state) Recover original ITER capability for nuclear testing Would address several physics tasks requested by IT Leader Next Step Option Activities

Magnets and PFCs (power and particle-handling, including tritium inventory): –How disruptions/VDEs which may affect the ITER design. –Characterization of thermal energy load during disruption –Model development of halo current width during VDEs based on experiments –Simulations of VDEs in ITER with 3D MHD code –Disruption mitigation by noble gas injection –Oxygen baking experiment, which could be possible during spring 2005 at D III-D and is under discussion at GA, may be one of the possible tasks. Heating and Current-drive and advanced control: –ITER Plasma Integrated Model for ITER for Control –Feasibility study of ITER SS scenarios with high confinement, NBCD, ECCD, LHCD, ICCD and fueling by pellet injection. –RF launchers –Validation of enhanced confinement models and application to ITER. –Development of Steady State Scenarios in ITER –RWM in Steady State Scenario in ITER –Evaluation of Fast Particle Confinement of ITER Diagnostics: –Specific diagnostic design tasks, including updating procurement packages [activities related to the diagnostics for which the US is responsible] Physics Tasks Requested by the International Team Leader [US ITER Project TOFE, 9/04] (TAE instabilities in Steady-state scenarios)

First Results from ITER-AT Studies Using TSC,TRANSP and NOVA-K Goal is Steady-State,  N ≈ 3.5, f bs > 60% f bs, Q > 5 using NINB, ICFW and LHCD First case has  N ≈ 2.5, f bs ≈ 44%, ≈ 97% non-inductive and Q ≈ 5. Optimize CD mix & startup to flatten q profile FT/P7-23

ITER PF Coils Baseline RWM coils located outside TF coils Applying FIRE-Like RWM Feedback Coils to ITER Increases  limit for n = 1 from  N = 2.5 to 4.9 Engineering feasibility of internal control coils needs to be determined VALEN Analysis Columbia University FIRE-like RWM coils would have large stabilizing effect on n=1 No-wall limit RWM Coils in every third port G. Navratil, J. Bialek FIRE-like RWM coils would be located on port shield plugs inside the vacuum vessel.

The proposed RWM Coils would be in the Front Assembly of Every 3rd Port Plug Assembly RWM coils wrapped on end of Port Plug

TAE Modes in ITER Hybrid Regime   = 2.1%, R   =15.5 %, T e (0) = 34 keV n = unstable (only odd modes have been studied). radially localized near r/a = 0.5 TAEs are strongly driven with beam ions as much as alphas (H.Berk talk) strong drive is due to high   as a result of high ion temperature Perturbative approach may not be valid. Strong particle transport is expected (H.Berk talk). Nova-K: N. Gorelenkov

FIRE Pre-Conceptual Design has been completed - exceeding original goals. Steady-state FIRE AT mode using FWCD and LHCD has f bs ≈ 77% and is 100% non-inductive. High  N (~4) with close coupled RWM stabilization would produce power-plant-level fusion-power densities of 5 MWm -3 for 4  CR. FIRE has passed DOE PVR and is ready to be put forward as an attractive burning plasma experiment if the six-party ITER negotiations breakdown. Transition of NSO activities to supporting the “option” of extending ITER performance using AT operation has been accomplished without missing a step. The goal is to recover most of the capability of the original ITER. Initial ITER-AT studies successfully at producing steady-state scenario using FWCD and LHCD with f bs ≈ 48% and 97% non-inductive current drive. Studies are continuing to optimize the current drive, and to increase  N from 2.5 to 3.5. Close coupled RWM coils proposed by FIRE are expected to provide n = 1 stability up to  N = 4.2 in FIRE and 4.9 in ITER. TAE instability studies indicate that AT modes in both ITER and FIRE will have multiple unstable modes, with NINB ions being a significant drive term in ITER. Concluding Remarks

I want you to get on with fusion, Uncle Sam’s Thoughts on Fusion and don’t you dare go back to Square 1!