Plasma Stability in Tokamaks and Stellarators
|
|
- Lenard Hodge
- 3 years ago
- Views:
Transcription
1 Plasma Stability in Tokamaks and Stellarators Gerald A. Navratil GCEP Fusion Energy Workshop Princeton, NJ 1- May 006
2 ACKNOWLEDGEMENTS Borrowed VGs from many colleagues: J. Bialek, A. Garofalo,R. Goldston, A. Hubbard, R. Lahaye, J. Menard, H. Neilson, M. Okabayashi, E. J. Strait, S. Sabbagh, T. Taylor, M. Zarnstorff,
3 MHD EQUILIBRIUM AND STABILITY MHD Equilibrium requires: p = J x B MHD sets β limit Loss of equilibrium Sources of free-energy for instability: Magnetic: B /µ o Pressure: nt Three primary limiting β phenomena: Long Wavelength Ideal Modes: n = 0, 1,, 3 Short Wavelength Ideal Modes: n Long Wavelength Resistive Modes: n = 1, Magnetic Reconnection: E + v x B = ηj
4 β LIMITING MODES: LOW-n Must deal with long wavelength modes Shift & Tilt: n = 0 and 1 Kink: n = 1 n=0 mode: in tokamaks is solved with wall stabilization & active feedback control n=1 kink mode: limits determined to 10% to 0%. Tokamaks: solutions in hand using plasma rotation & active control. Stellarators: external magnetic field transform may allow sufficiently high beta below kink limit.
5 β LIMITING MODES: HIGH-n Must deal with short wavelength ballooning/interchange modes, n Limits well understood and with 10% to 0% accuracy compared with experiments Controlled by plasma shape and magnetic shear profile:
6 β LIMITING MODES: TEARING MODES Must deal with long wavelength resistive tearing modes, n = 1,, Magnetic field line puncture plot showing island structure Tearing Modes: In tokamaks: stabilized by current profile control and active control with local ECH. In stellarators: controlled by tailoring external transform.
7 A COMPACT STEADY STATE TOKAMAK REQUIRES OPERATION AT HIGH β N P fus γ ε cur Q ss = eff β N B 3 aκ P CD nq ( 1 ξ A q β ) N β Power Density T ε 1 + κ DIII D NATIONAL FUSION FACILITY SAN DIEGO Current Limit q* = 4 Advanced Pressure Conventional Tokamak Limit Tokamak β N = 5 β N = 3.5 Equilibrium Limit εβ p Bootstrap Current High power density high β T Large bootstrap fraction high β p Steady state high β N β N power density bootstrap current ( 1 + κ β T β p )β N β N = β T /(I/aB) 130 0/TST/wj
8 PRIMARY LIMITING MODE IN MAGNETIC CONFINEMENT SYSTEMS: LOW-n Kink Long wavelength global MHD modes driven by pressure & current gradient: Shift & Tilt: n = 0 and 1 Kink: n = 1 Classic Instability: Ideal conducting wall on plasma boundary stabilizes the kink mode by freezing magnetic flux value on wall surface. Resistive conducting wall stabilization fails on magnetic field soak-through time scale: τ w
9 perturbed magnetic energy δw = 1 3 d x {ε c δb + ε c ( B ) (ξ δb) p Foundation of Kink Mode Stability Built on Energy Principle δw Stability Analysis 1957 Bernstein, Frieman, Kruskal, Kulsrud o } + ( ξ)(ξ p o ) + γp ( ξ) o pressure driven - destabilizing 1 δw = d x ε c δb current driven - destabilizing o plasma compression 3 o v vacuum perturbed magnetic energy If δw + δw < 0 mode is unstable p v
10 BASIC KINK MODE Long wavelength mode driven by pressure & current gradient Cylindrical k ~ π/l Toroidal: low n = 1 Unstable when δw p + δw v < 0 Dispersion Relation: γ K + δw p + δw v = 0, where K is kinetic fluid mass Define Γ = [δw p + δw v]/k ~ [v Alfvén /L]
11 IDEAL WALL STABILIZES THE KINK MODE Ideal wall traps field in vacuum region and restoring force stabilizes the kink EXTERNAL Kink: Unstable when δw p + δw d v < 0 Note: δw d v > δw v Dispersion Relation: γ - Γ + [δw d v-δw v]/k = 0 Critical Wall Distance, d c, where kink stable for d < d c : simple [δw d v-δw v]/k parameterization with d: γ - Γ [1 d c /d]/k = 0
12 KINK MODE IS STABILIZED BY IDEAL WALL 0 = γ Γ (1 d c) } Ideal Stability d γ / Γ ideal mode stable Ideal Instability γ Γ ideal mode unstable Plasma-Wall Separation, d/d c /GAN/rs
13 RESISTIVE WALL LEAKS STABILIZING FIELD: τ W Stabilizing field decays resistively on wall time scale τ w ~ L/R: dψ w /dt = - ψ w /τ w Quadratic kink: γ - Γ [1-d c /d] = 0 coupled to slow flux diffusion γψ w = - ψ w /τ w : τ w >> τ Alfvén Cubic Dispersion Relation with new slow root the RWM: γ - Γ [1-(d c /d) γτ w /(γτ w + 1)] = 0
14 KINK MODE GROWTH IS SLOWED BY RESISTIVE WALL 0 = ( γ τ γ Γ 1 d w ) c d γ τ w + 1 } Ideal Stability } Resistive Wall γ / Γ Real ω 0 γ τ w 1 Resistive Wall Mode ideal mode unstable Plasma-Wall Separation, d/d c Resistive wall mode (RWM) is unstable Mode structure similar to ideal external kink Mode grows slowly: γ ~ τ 1 w /GAN/rs
15 RWM STABILIZED IN DIII-D BY ROTATION FOR MANY WALL-TIMES, τ W Normalized plasma pressure, β N, exceeds no-wall stability limit by up to 40% n = 1 mode grows (γ ~ 1/τ W ) after toroidal rotation at q = 3 surface has decreased below ~1 khz /GAN/rs
16 ROTATION AND DISSIPATION CAN STABILIZE RWM Rotation Doppler shift: γ γ + iω where Ω is plasma rotation. Dissipation represented by friction loss (γ + iω)ν, where form of ν still being actively studied by theory community: (γ + iω) - Γ [1-(d c /d) γτ w /(γτ w + 1)] + (γ + iω)ν = 0 (as shown in Chu, et al. Phys. Plasma 1995; consistent with numerical result of Bondeson & Ward, PRL 1994) Cubic Dispersion Relation with three roots: in region where d < d c new slow RWM root can be damped with fast stable kink mode roots tied to rotating plasma with usual ordering: τ w -1 << Ω << v Alfvén /L Why is RWM Slow Root Stabilized? kink energy release < dissipation loss of RWM slowed by wall in flowing plasma
17 γ / Γ KINK MODE GROWTH IS SLOWED BY RESISTIVE WALL AND STABILIZED BY PLASMA ROTATION (γ + iω) Γ 0 = + } Ideal Stability Stable Gap ω τ w 1 Resistive Wall Mode Plasma Mode γ Γ Plasma-Wall Separation, d/d c Γ (d c /d)γ τ w γ τ w + 1 } Resistive Wall + (γ + iω) ν DIS } Plasma Dissipation Resistive wall mode (RWM) is unstable Mode structure similar to ideal external kink Mode grows slowly: γ ~ τ 1 w Dissipation + rotation stabilizes RWM Mode nearly stationary: ω ~ τw 1 << Ω plasma /GAN/rs
18 SUSTAINED ROTATION ABOVE CRITICAL VALUE RELIABLE OPERATION ABOVE THE NO-WALL LIMIT β N Feedback control of NBI power keeps β N below stability limit (107603) 1 no wall β N (.4l i) 0 1 Rotation (khz) at q~ Time (ms) No other large scale instabilities encountered (NTM, n= RWM,... ) Ideal n=1 kink observed at the wall-stabilized β limit Toroidal Angle δb p β N ~ β no-wall N β = 3.7% τ g ~ 300 µs << τ wall DIII D NATIONAL FUSION FACILITY SAN DIEGO Time (ms) T rot ~1 ms < τ wall 58 0/EJS/wj
19 MARS PREDICTIONS OF Ω crit τ A IN QUALITATIVE AGREEMENT WITH MEASUREMENTS ON DIII-D AND JET sound wave In DIII-D Ω crit τ A ~ 0.0 with weak β dependence In JET Ω crit τ A ~ with weak β dependence Both damping models predict Ω crit within a factor of
20 MODE FREQUENCY AND DAMPING CANNOT BE FIT SIMULTANEOUSLY Growth rate γ RWM τ W experiment(ωτ ~0.0) A kinetic sound wave (κ = 0.5) Mode rotation frequency ω RWM τ W Both damping models predict γ too low RWM Kinetic damping predicts mode frequency ω RWM Further work on damping [e.g. neoclassical viscosity] models being explored DIII D NATIONAL FUSION FACILITY S A N D I E G O C β
21 NSTX provides crucial data for understanding the dissipation mechanisms that allow rotational stabilization of the RWM Insight from drift-kinetic theory: Trapped-particle effects at finite ε significantly weaken ion Landau damping, but Toroidal inertia enhancement modifies eigenfunction when Ω φ / ω A > 1/4q (Columbia Univ.) Experimental Ω crit / ω A suggests scaling ε / q why? Is dissipation localized to resonant surfaces, or more global? Addressing questions above w/ NSTX / DIII-D similarity experiments, and hi-res CHERS ST has uniquely high ω sound / ω A distinguish between ω s and ω A scaling NSTX DIII-D shifted & scaled 1.1 Needed for predicting control requirements for RWM stabilization in ITER & CTF 13
22 FEEDBACK LOGIC FOR RWM FEEDBACK STABILIZATION Smart Shell Explicit Mode Control Feedback cancels the radial flux from MHD mode at wall sensor Feedback cancels the flux from MHD mode at plasma surface /GAN/rs
23 DIII D INTERNAL CONTROL COILS ARE PREDICTED TO PROVIDE STABILITY AT HIGHER BETA Inside vacuum vessel: Faster time response for feedback control Closer to plasma: more efficient coupling Internal Coils (I-coils) /GAN/rs
24 FEEDBACK WITH I-COILS IN DIII-D INCREASES STABLE PLASMA PRESSURE TO NEAR IDEAL-WALL LIMIT VALEN code prediction Normalized Growth Rate γτ w No Feedback Ideal kink VALEN code: - DCON MHD stability - 3D geometry of vacuum vessel and coil geometry Resistive Wall Mode: Open loop growth rate τ w is the vacuum vessel flux diffusion time (~ 3.5 ms) 0 0. No-Wall Limit 0.4 C β Ideal-Wall Limit DIII D NATIONAL FUSION FACILITY S A N D I E G O 69-04/MO/jy
25 FEEDBACK WITH I-COILS IN DIII-D INCREASES STABLE PLASMA PRESSURE TO NEAR IDEAL-WALL LIMIT C-coil stabilizes slowly growing RWMs Normalized Growth Rate γτ w No-Wall Limit No Feedback External C-coils Accessible with External C-coils 0.4 C β 0.6 Ideal kink Ideal-Wall Limit External C-Coil: - Control fields must penetrate wall - Induced eddy currents reduce feedback τ w is the vacuum vessel flux diffusion time (~3.5 ms) DIII D NATIONAL FUSION FACILITY S A N D I E G O 69-04/MO/jy
26 I-coil stabilizes RWMs with growth rate 10 times faster than C-coils Normalized Growth Rate γτ w FEEDBACK WITH I-COILS IN DIII-D INCREASES STABLE PLASMA PRESSURE TO NEAR IDEAL-WALL LIMIT No-Wall Limit No Feedback External C-coils Internal I-coils Accessible with External C-coils 0.4 C β 0.6 Accessible with Internal I-coils Ideal kink Ideal-Wall Limit Internal I-Coils: - Improved coil/plasma coupling - Improved spatial match to RWM field structure τ w is the vacuum vessel flux diffusion time (~3.5 ms) DIII D NATIONAL FUSION FACILITY S A N D I E G O 69-04/MO/jy
27 FEEDBACK EFFICACY DEMONSTRATED BY GATING OFF THE GAIN FOR 0 MS AT TIME OF EXPECTED RWM ONSET (gauss) Feedback Gain β N n=1 δb r Time (ms) Without feedback, slow Ip ramp rate (0.5 MA/s) destabilizes slowly growing RWM With feedback, beta collapse avoided (cm) 1 0 Relative Displacement (SXR) n=1 mode starts up during feedback off period, stabilized after feedback is turned back on (ka) Feedback Current (C79) Feedback OFF 1530 Time (ms) n=1 mode detected on poloidal field probes and SXR arrays, decoupled from driver coils DIII D NATIONAL FUSION FACILITY S A N D I E G O
28 FEEDBACK WITH INTERNAL CONTROL COILS HAS ACHIEVED HIGH C β AT ROTATION BELOW CRITICAL LEVEL PREDICTED BY MARS Trajectories of plasma discharge in rotation versus C β No feedback plasma approaches limit and disrupts ideal wall limit C β C-coil feedback plasma crosses limit & reaches higher pressure no wall limit 1.0 Unstable (without feedback) C-COIL MARS prediction Stable (without feedback) NO FEEDBACK time Rotation (km/s) DIII D NATIONAL FUSION FACILITY S A N D I E G O 69-04/MO/jy
29 FEEDBACK WITH INTERNAL CONTROL COILS HAS ACHIEVED HIGH C β AT ROTATION BELOW CRITICAL LEVEL PREDICTED BY MARS With near zero Rotation, Cβ is near the maximum set by existing control system characteristics: bandwidth & processing time delay ideal wall 1.0 limit I-coil feedback plasma reaches near zero rotation C β Unstable MARS prediction Stable (without feedback) MARS /VALEN prediction with measured amplifier time response for zero rotation no wall limit I-COIL ZERO ROTATION C-COIL NO FEEDBACK time Rotation (km/s) DIII D NATIONAL FUSION FACILITY S A N D I E G O 69-04/MO/jy
30 FEEDBACK WITH INTERNAL CONTROL COILS HAS ACHIEVED HIGH C β AT ROTATION BELOW CRITICAL LEVEL PREDICTED BY MARS Combination of low rotation and feedback reaches C β is the ideal wall-limits ideal wall limit 1.0 Unstable I-COIL MARS prediction Stable (without feedback) C β MARS /VALEN prediction with measured power supply time response for zero rotation no wall limit I-COIL ZERO ROTATION C-COIL NO FEEDBACK time Rotation (km/s) DIII D NATIONAL FUSION FACILITY S A N D I E G O 69-04/MO/jy
31 RWM FEEDBACK ASSISTS IN EXTENDING β n ~4 ADVANCED TOKAMAK DISCHARGE MORE THAN 1 SECOND 4.0 βn Feedback current (ka) Plasma Rotation (km/s) n=1 δb p Mode Amplitude (gauss) 0.0 F.B on No Feedback No Feedback With Feedback Estimated no-wall limit / Feedback coil current amplitude With Feedback Time (ms) DIII D NATIONAL FUSION FACILITY S A N D I E G O High performance plasma approaches β ~ 6% Without feedback plasma disrupts due to RWM 69-04/MO/jy
32 RWM Stabilization Has Opened New High Performance Regimes Above the No-Wall Stability Limit Simultaneous feedback control of error fields and RWM Additional ECCD power in FY06 will help sustain high q min New divertor will help density control at high triangularity β T (%) β N 6l i 4l i Time (ms) Pressure (10 5 Pa) J (A/cm ) Safety Factor ρ DIII D NATIONAL FUSION FACILITY /EJS/rs
33 Applying Internal RWM Feedback Coils to the Port Plugs in ITER Increases β limit for n = 1 from β N =.5 to ~ 4 RWM Coil Concept for ITER VALEN Analysis Columbia University Baseline RWM coils located outside TF coils Internal RWM coils would be located inside No-wall limit the vacuum vessel behind shield module 7 RWM Coils mounted behind the BSM in but inside the vacuum vessel on the every other port except NBI ports. removable port plugs. (assumes 9 ms time constant for each BSM) Integration and Engineering feasibility of internal RWM coils is under study.
34 NCSX Compact Stellarator Low-n Stability Stellarators provide external magnetic field transform aiming at: Steady state without current drive. Kink stable at sufficiently high pressure (β > 4%) without feedback control or rotation drive. Compact Stellarators (CS) improve on previous designs. Magnetic quasi-symmetry: good confinement. link to tokamak physics. Lower aspect ratio. 3-Period NCSX Plasma and Coil Design
35 NCSX Stability Modeling Predicts Kink Stability up to 6% PIES Free-Boundary Equilibrium at β = 4.1%
36 W7AS: β 3.4 % : Quiescent, Quasi-stationary (100 m-3) <!> (%) Power (MW) (A.U.) <!> 540 ne Mirnov Ḃ P NB 1 P rad Time (s) MCZ B = 0.9 T, iota vac 0.5 Almost quiescent high- β phase, MHD-activity in early medium-β phase In general, β not limited by any detected MHD-activity. I P = 0, but there can be local currents Similar to High Density H-mode (HDH) Similar β>3.4% plasmas achieved with B = T with either NBI-alone, or combined NBI + OXB ECH heating. Much higher than predicted β limit ~ %
37 Current-carrying Systems are Subject to Reconnection Tearing Modes Normal magnetic shear Safety Factor 6 4 MSE data 3/ Reverse magnetic shear Minor radius (r/a) 1.0 A rational field line can be an O point around which islands form. (- j for normal shear, + j for reverse shear)
38 Pressure-driven Bootstrap Current is a Boon and a Bane - In the presence of a pressure gradient, trapped particles entrain a parallel bootstrap current. A neoclassical effect, i.e. collisional, but including nonlocal orbit effects. - May allow steady-state operation of axisymmetric toroidal systems. - Drives neoclassical tearing modes in normal shear regions due to current depletion in the magnetic islands. µ 0 dw 1.! nc dt = #" + a L 1$ 1 q % & L p ' ) ( w w + w c *., - a &i % & g($) ' + w 3 ) ( L q L p *, + Ohmic current Bootstrap current Polarization current Finite transport correction
39 Theory Accurately Predicts Growth of Neoclassical Tearing Modes (NTM) W (cm) 5 R = 3m, T e ~ 5keV Magnetic 4 3 Theory ECE 1 4MW NBI Time (s) w (cm) R = 1m, T e ~ 1keV time (s)! p(meas) Shot Measured Island Width (cm) 0. 0 Island Width Predicted by Neoclassical Theory (cm) ! p(meas) Bootstrap current + normal shear drives NTM s. - Agrees to factor of ~ with neoclassical resistivity, over a wide range of plasma parameters. - Important challenge to theory of magnetic reconnection. - Reverse shear stabilizes NTM s, as predicted. - Strong implications for toroidal system optimization.
40 Replacing Bootstrap Current in Islands Stabilizes Neoclassical Tearing Modes ECCD Steerable Electron Cyclotron Current Drive wave launcher. ITER will have ECCD for NTM control.
41 DIII D Demonstrates NTM Active Stabilization with ECCD P NB (MW) Current 10 (MA) P NB (MW) ~ B (n = 1) (G) EC Power (MW) High beta is achieved with preemptive stabilization of the /1 NTM Stable operation at the no-wall beta limit for >1 s Div. D α (au) 4 l i β N ECCD is applied before the mode appears Real-time tracking of the q= surface maintains current drive alignment Tearing mode appears promptly when ECCD is removed B (T) Time (ms) DIII D NATIONAL FUSION FACILITY
42 ECCD in ITER Can Reduce the m/n=/1 NTM Island Cross-machine bench marking... R.J. La Haye, et.al., submitted to Nuclear Fusion Locking condition from 0-D model w 3 w ω 0 τ A0 a (1 + 0 a)= * 14 1 τ w τ E0 Island growth rate (τ R /r) dw/dt ITER, m/n=/1, β N = 1.84 NO ECCD Unstable Region 1 MW ECCD No Modulation (K 1 = 0.38, F=1) NO ECCD Saturated Island (if Beta Maintained and if Mode Does Not Lock) m/n=/1 Island full width w (cm) 1 MW ECCD 50/50 Modulation (K 1 = 0.74, F = 0.5) DIII D ITER...τ w = ms (J. Bialek)...f 0 = khz (A. Polevoi)...τ E0 = s (J. Cordey)...τ A0 = μs (Y. Gribov)... w = a lock w 5 cm in ITER to lock w 10 cm at f 0 = 1.4 khz 53-05/RJL/jy
43 Key Open Issues in Stability Advanced Tokamak Spherical Torus Low-n Kink: Rotation Stabilization Physics & Scaling Scale Active Feedback to ITER - n = 1,, Coil Modularity & Failure of Mode Rigidity Low-n Tearing: NTM Active Control Requirements for ITER Quantitative Theory: Seeding physics, island rotation, ECCD localization & modulation, small island modeling Compact Stellarator Low-n Kink: Validate no-wall high-beta kink limits Rotation effects in QS equilibria Low-n Tearing: NTM control with external magnetic transform + large bootstrap current Magnetic Island control as β and Ip vary.
Resistive Wall Mode Control in DIII-D
Resistive Wall Mode Control in DIII-D by Andrea M. Garofalo 1 for G.L. Jackson 2, R.J. La Haye 2, M. Okabayashi 3, H. Reimerdes 1, E.J. Strait 2, R.J. Groebner 2, Y. In 4, M.J. Lanctot 1, G.A. Navratil
RESISTIVE WALL MODE STABILIZATION RESEARCH ON DIII D STATUS AND RECENT RESULTS
RESISTIVE WALL MODE STABILIZATION RESEARCH ON STATUS AND RECENT RESULTS by A.M. Garofalo1 in collaboration with J. Bialek,1 M.S. Chance,2 M.S. Chu,3 T.H. Jensen,3 L.C. Johnson,2 R.J. La Haye,3 G.A. Navratil,1
DIII D. by M. Okabayashi. Presented at 20th IAEA Fusion Energy Conference Vilamoura, Portugal November 1st - 6th, 2004.
Control of the Resistive Wall Mode with Internal Coils in the Tokamak (EX/3-1Ra) Active Measurement of Resistive Wall Mode Stability in Rotating High Beta Plasmas (EX/3-1Rb) by M. Okabayashi Presented
RWM Control in FIRE and ITER
RWM Control in FIRE and ITER Gerald A. Navratil with Jim Bialek, Allen Boozer & Oksana Katsuro-Hopkins MHD Mode Control Workshop University of Texas-Austin 3-5 November, 2003 OUTLINE REVIEW OF VALEN MODEL
Active MHD Control Needs in Helical Configurations
Active MHD Control Needs in Helical Configurations M.C. Zarnstorff 1 Presented by E. Fredrickson 1 With thanks to A. Weller 2, J. Geiger 2, A. Reiman 1, and the W7-AS Team and NBI-Group. 1 Princeton Plasma
RWM FEEDBACK STABILIZATION IN DIII D: EXPERIMENT-THEORY COMPARISONS AND IMPLICATIONS FOR ITER
GA A24759 RWM FEEDBACK STABILIZATION IN DIII D: EXPERIMENT-THEORY COMPARISONS AND IMPLICATIONS FOR ITER by A.M. GAROFALO, J. BIALEK, M.S. CHANCE, M.S. CHU, D.H. EDGELL, G.L. JACKSON, T.H. JENSEN, R.J.
Effect of Resonant and Non-resonant Magnetic Braking on Error Field Tolerance in High Beta Plasmas
1 EX/5-3Ra Effect of Resonant and Non-resonant Magnetic Braking on Error Field Tolerance in High Beta Plasmas H. Reimerdes 1), A.M. Garofalo 2), E.J. Strait 2), R.J. Buttery 3), M.S. Chu 2), Y. In 4),
Performance limits. Ben Dudson. 24 th February Department of Physics, University of York, Heslington, York YO10 5DD, UK
Performance limits Ben Dudson Department of Physics, University of York, Heslington, York YO10 5DD, UK 24 th February 2014 Ben Dudson Magnetic Confinement Fusion (1 of 24) Previously... In the last few
The Effects of Noise and Time Delay on RWM Feedback System Performance
The Effects of Noise and Time Delay on RWM Feedback System Performance O. Katsuro-Hopkins, J. Bialek, G. Navratil (Department of Applied Physics and Applied Mathematics, Columbia University, New York,
DIAGNOSTICS FOR ADVANCED TOKAMAK RESEARCH
DIAGNOSTICS FOR ADVANCED TOKAMAK RESEARCH by K.H. Burrell Presented at High Temperature Plasma Diagnostics 2 Conference Tucson, Arizona June 19 22, 2 134 /KHB/wj ROLE OF DIAGNOSTICS IN ADVANCED TOKAMAK
NIMROD FROM THE CUSTOMER S PERSPECTIVE MING CHU. General Atomics. Nimrod Project Review Meeting July 21 22, 1997
NIMROD FROM THE CUSTOMER S PERSPECTIVE MING CHU General Atomics Nimrod Project Review Meeting July 21 22, 1997 Work supported by the U.S. Department of Energy under Grant DE-FG03-95ER54309 and Contract
Analysis and modelling of MHD instabilities in DIII-D plasmas for the ITER mission
Analysis and modelling of MHD instabilities in DIII-D plasmas for the ITER mission by F. Turco 1 with J.M. Hanson 1, A.D. Turnbull 2, G.A. Navratil 1, C. Paz-Soldan 2, F. Carpanese 3, C.C. Petty 2, T.C.
Dependence of Achievable β N on Discharge Shape and Edge Safety Factor in DIII D Steady-State Scenario Discharges
Dependence of Achievable β N on Discharge Shape and Edge Safety Factor in DIII D Steady-State Scenario Discharges by J.R. Ferron with T.C. Luce, P.A. Politzer, R. Jayakumar, * and M.R. Wade *Lawrence Livermore
(a) (b) (c) (d) (e) (f) r (minor radius) time. time. Soft X-ray. T_e contours (ECE) r (minor radius) time time
Studies of Spherical Tori, Stellarators and Anisotropic Pressure with M3D 1 L.E. Sugiyama 1), W. Park 2), H.R. Strauss 3), S.R. Hudson 2), D. Stutman 4), X-Z. Tang 2) 1) Massachusetts Institute of Technology,
Extended Lumped Parameter Model of Resistive Wall Mode and The Effective Self-Inductance
Extended Lumped Parameter Model of Resistive Wall Mode and The Effective Self-Inductance M.Okabayashi, M. Chance, M. Chu* and R. Hatcher A. Garofalo**, R. La Haye*, H. Remeirdes**, T. Scoville*, and T.
Advances in Global MHD Mode Stabilization Research on NSTX
1 EX/5-1 Advances in Global MHD Mode Stabilization Research on NSTX S.A. Sabbagh 1), J.W. Berkery 1), R.E. Bell 2), J.M. Bialek 1), S.P. Gerhardt 2), J.E. Menard 2), R. Betti 3), D.A. Gates 2), B. Hu 3),
Disruption dynamics in NSTX. long-pulse discharges. Presented by J.E. Menard, PPPL. for the NSTX Research Team
Disruption dynamics in NSTX long-pulse discharges Presented by J.E. Menard, PPPL for the NSTX Research Team Workshop on Active Control of MHD Stability: Extension of Performance Monday, November 18, 2002
Effects of Noise in Time Dependent RWM Feedback Simulations
Effects of Noise in Time Dependent RWM Feedback Simulations O. Katsuro-Hopkins, J. Bialek, G. Navratil (Department of Applied Physics and Applied Mathematics, Columbia University, New York, NY USA) Building
Global Mode Control and Stabilization for Disruption Avoidance in High-β NSTX Plasmas *
1 EX/P8-07 Global Mode Control and Stabilization for Disruption Avoidance in High-β NSTX Plasmas * J.W. Berkery 1, S.A. Sabbagh 1, A. Balbaky 1, R.E. Bell 2, R. Betti 3, J.M. Bialek 1, A. Diallo 2, D.A.
MHD. Jeff Freidberg MIT
MHD Jeff Freidberg MIT 1 What is MHD MHD stands for magnetohydrodynamics MHD is a simple, self-consistent fluid description of a fusion plasma Its main application involves the macroscopic equilibrium
Advanced Tokamak Research in JT-60U and JT-60SA
I-07 Advanced Tokamak Research in and JT-60SA A. Isayama for the JT-60 team 18th International Toki Conference (ITC18) December 9-12, 2008 Ceratopia Toki, Toki Gifu JAPAN Contents Advanced tokamak development
Formation and Long Term Evolution of an Externally Driven Magnetic Island in Rotating Plasmas )
Formation and Long Term Evolution of an Externally Driven Magnetic Island in Rotating Plasmas ) Yasutomo ISHII and Andrei SMOLYAKOV 1) Japan Atomic Energy Agency, Ibaraki 311-0102, Japan 1) University
STABILIZATION OF THE RESISTIVE WALL MODE IN DIII D BY PLASMA ROTATION AND MAGNETIC FEEDBACK
GA A24014 STABILIZATION OF THE RESISTIVE WALL MODE IN DIII D BY PLASMA ROTATION AND MAGNETIC FEEDBACK by M. Okabayashi, J. Bialek, M.S. Chance, M.S. Chu, E.D. Fredrickson, A.M. Garofalo, R. Hatcher, T.H.
GA A26247 EFFECT OF RESONANT AND NONRESONANT MAGNETIC BRAKING ON ERROR FIELD TOLERANCE IN HIGH BETA PLASMAS
GA A26247 EFFECT OF RESONANT AND NONRESONANT MAGNETIC BRAKING ON ERROR FIELD TOLERANCE IN HIGH BETA PLASMAS by H. REIMERDES, A.M. GAROFALO, E.J. STRAIT, R.J. BUTTERY, M.S. CHU, Y. In, G.L. JACKSON, R.J.
STABILIZATION OF m=2/n=1 TEARING MODES BY ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII D TOKAMAK
GA A24738 STABILIZATION OF m=2/n=1 TEARING MODES BY ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII D TOKAMAK by T.C. LUCE, C.C. PETTY, D.A. HUMPHREYS, R.J. LA HAYE, and R. PRATER JULY 24 DISCLAIMER This
Dynamical plasma response of resistive wall modes to changing external magnetic perturbations
Dynamical plasma response of resistive wall modes to changing external magnetic perturbations M. Shilov, C. Cates, R. James, A. Klein, O. Katsuro-Hopkins, Y. Liu, M. E. Mauel, D. A. Maurer, G. A. Navratil,
GA A26242 COMPREHENSIVE CONTROL OF RESISTIVE WALL MODES IN DIII-D ADVANCED TOKAMAK PLASMAS
GA A26242 COMPREHENSIVE CONTROL OF RESISTIVE WALL MODES IN DIII-D ADVANCED TOKAMAK PLASMAS by M. OKABAYASHI, I.N. BOGATU, T. BOLZONELLA, M.S. CHANCE, M.S. CHU, A.M. GAROFALO, R. HATCHER, Y. IN, G.L. JACKSON,
Resistive Wall Mode Stabilization and Plasma Rotation Damping Considerations for Maintaining High Beta Plasma Discharges in NSTX
1 EXS/5-5 Resistive Wall Mode Stabilization and Plasma Rotation Damping Considerations for Maintaining High Beta Plasma Discharges in NSTX S.A. Sabbagh 1), J.W. Berkery 1), J.M. Bialek 1), R.E. Bell ),
1999 RESEARCH SUMMARY
1999 RESEARCH SUMMARY by S.L. Allen Presented to DIII D Program Advisory Committee Meeting January 2 21, 2 DIII D NATIONAL FUSION FACILITY SAN DIEGO 3 /SLA/wj Overview of Physics Results from the 1999
THE DIII D PROGRAM THREE-YEAR PLAN
THE PROGRAM THREE-YEAR PLAN by T.S. Taylor Presented to Program Advisory Committee Meeting January 2 21, 2 3 /TST/wj PURPOSE OF TALK Show that the program plan is appropriate to meet the goals and is well-aligned
KSTAR Equilibrium Operating Space and Projected Stabilization at High Normalized Beta
1 THS/P2-05 KSTAR Equilibrium Operating Space and Projected Stabilization at High Normalized Beta Y.S. Park 1), S.A. Sabbagh 1), J.W. Berkery 1), J.M. Bialek 1), Y.M. Jeon 2), S.H. Hahn 2), N. Eidietis
Dynamical plasma response of resistive wall modes to changing external magnetic perturbations a
PHYSICS OF PLASMAS VOLUME 11, NUMBER 5 MAY 2004 Dynamical plasma response of resistive wall modes to changing external magnetic perturbations a M. Shilov, b) C. Cates, R. James, A. Klein, O. Katsuro-Hopkins,
Characterization and Forecasting of Unstable Resistive Wall Modes in NSTX and NSTX-U *
1 EX/P4-34 Characterization and Forecasting of Unstable Resistive Wall Modes in NSTX and NSTX-U * J.W. Berkery 1, S.A. Sabbagh 1, Y.S. Park 1, R.E. Bell 2, S.P. Gerhardt 2, C.E. Myers 2 1 Department of
Effects of stellarator transform on sawtooth oscillations in CTH. Jeffrey Herfindal
Effects of stellarator transform on sawtooth oscillations in CTH Jeffrey Herfindal D.A. Ennis, J.D. Hanson, G.J. Hartwell, E.C. Howell, C.A. Johnson, S.F. Knowlton, X. Ma, D.A. Maurer, M.D. Pandya, N.A.
Three Dimensional Effects in Tokamaks How Tokamaks Can Benefit From Stellarator Research
1 TH/P9-10 Three Dimensional Effects in Tokamaks How Tokamaks Can Benefit From Stellarator Research S. Günter, M. Garcia-Munoz, K. Lackner, Ph. Lauber, P. Merkel, M. Sempf, E. Strumberger, D. Tekle and
- Effect of Stochastic Field and Resonant Magnetic Perturbation on Global MHD Fluctuation -
15TH WORKSHOP ON MHD STABILITY CONTROL: "US-Japan Workshop on 3D Magnetic Field Effects in MHD Control" U. Wisconsin, Madison, Nov 15-17, 17, 2010 LHD experiments relevant to Tokamak MHD control - Effect
Stationary, High Bootstrap Fraction Plasmas in DIII-D Without Inductive Current Control
Stationary, High Bootstrap Fraction Plasmas in DIII-D Without Inductive Current Control P. A. Politzer, 1 A. W. Hyatt, 1 T. C. Luce, 1 F. W. Perkins, 4 R. Prater, 1 A. D. Turnbull, 1 D. P. Brennan, 5 J.
Requirements for Active Resistive Wall Mode (RWM) Feedback Control
Requirements for Active Resistive Wall Mode (RWM) Feedback Control Yongkyoon In 1 In collaboration with M.S. Chu 2, G.L. Jackson 2, J.S. Kim 1, R.J. La Haye 2, Y.Q. Liu 3, L. Marrelli 4, M. Okabayashi
DIII D Research in Support of ITER
Research in Support of ITER by E.J. Strait and the Team Presented at 22nd IAEA Fusion Energy Conference Geneva, Switzerland October 13-18, 28 DIII-D Research Has Made Significant Contributions in the Design
DIII D. by F. Turco 1. New York, January 23 rd, 2015
Modelling and Experimenting with ITER: the MHD Challenge by F. Turco 1 with J.M. Hanson 1, A.D. Turnbull 2, G.A. Navratil 1, F. Carpanese 3, C. Paz-Soldan 2, C.C. Petty 2, T.C. Luce 2, W.M. Solomon 4,
A New Resistive Response to 3-D Fields in Low Rotation H-modes
in Low Rotation H-modes by Richard Buttery 1 with Rob La Haye 1, Yueqiang Liu 2, Bob Pinsker 1, Jong-kyu Park 3, Holger Reimerdes 4, Ted Strait 1, and the DIII-D research team. 1 General Atomics, USA 2
Localized Electron Cyclotron Current Drive in DIII D: Experiment and Theory
Localized Electron Cyclotron Current Drive in : Experiment and Theory by Y.R. Lin-Liu for C.C. Petty, T.C. Luce, R.W. Harvey,* L.L. Lao, P.A. Politzer, J. Lohr, M.A. Makowski, H.E. St John, A.D. Turnbull,
INTRODUCTION TO BURNING PLASMA PHYSICS
INTRODUCTION TO BURNING PLASMA PHYSICS Gerald A. Navratil Columbia University American Physical Society - Division of Plasma Physics 2001 Annual Meeting, Long Beach, CA 1 November 2001 THANKS TO MANY PEOPLE
Control of Sawtooth Oscillation Dynamics using Externally Applied Stellarator Transform. Jeffrey Herfindal
Control of Sawtooth Oscillation Dynamics using Externally Applied Stellarator Transform Jeffrey Herfindal D.A. Ennis, J.D. Hanson, G.J. Hartwell, S.F. Knowlton, X. Ma, D.A. Maurer, M.D. Pandya, N.A. Roberds,
GA A27444 PROBING RESISTIVE WALL MODE STABILITY USING OFF-AXIS NBI
GA A27444 PROBING RESISTIVE WALL MODE STABILITY USING OFF-AXIS NBI by J.M. HANSON, F. TURCO M.J. LANCTOT, J. BERKERY, I.T. CHAPMAN, R.J. LA HAYE, G.A. NAVRATIL, M. OKABAYASHI, H. REIMERDES, S.A. SABBAGH,
Der Stellarator Ein alternatives Einschlusskonzept für ein Fusionskraftwerk
Max-Planck-Institut für Plasmaphysik Der Stellarator Ein alternatives Einschlusskonzept für ein Fusionskraftwerk Robert Wolf robert.wolf@ipp.mpg.de www.ipp.mpg.de Contents Magnetic confinement The stellarator
Control of Neo-classical tearing mode (NTM) in advanced scenarios
FIRST CHENGDU THEORY FESTIVAL Control of Neo-classical tearing mode (NTM) in advanced scenarios Zheng-Xiong Wang Dalian University of Technology (DLUT) Dalian, China Chengdu, China, 28 Aug, 2018 Outline
Progressing Performance Tokamak Core Physics. Marco Wischmeier Max-Planck-Institut für Plasmaphysik Garching marco.wischmeier at ipp.mpg.
Progressing Performance Tokamak Core Physics Marco Wischmeier Max-Planck-Institut für Plasmaphysik 85748 Garching marco.wischmeier at ipp.mpg.de Joint ICTP-IAEA College on Advanced Plasma Physics, Triest,
Effect of an error field on the stability of the resistive wall mode
PHYSICS OF PLASMAS 14, 022505 2007 Effect of an error field on the stability of the resistive wall mode Richard Fitzpatrick Institute for Fusion Studies, Department of Physics, University of Texas at Austin,
Current-driven instabilities
Current-driven instabilities Ben Dudson Department of Physics, University of York, Heslington, York YO10 5DD, UK 21 st February 2014 Ben Dudson Magnetic Confinement Fusion (1 of 23) Previously In the last
Non-Solenoidal Plasma Startup in
Non-Solenoidal Plasma Startup in the A.C. Sontag for the Pegasus Research Team A.C. Sontag, 5th APS-DPP, Nov. 2, 28 1 Point-Source DC Helicity Injection Provides Viable Non-Solenoidal Startup Technique
(Motivation) Reactor tokamaks have to run without disruptions
Abstract A database has been developed to study the evolution, the nonlinear effects on equilibria, and the disruptivity of locked and quasi-stationary modes with poloidal and toroidal mode numbers m=2
Neoclassical Tearing Modes
Neoclassical Tearing Modes O. Sauter 1, H. Zohm 2 1 CRPP-EPFL, Lausanne, Switzerland 2 Max-Planck-Institut für Plasmaphysik, Garching, Germany Physics of ITER DPG Advanced Physics School 22-26 Sept, 2014,
Innovative Concepts Workshop Austin, Texas February 13-15, 2006
Don Spong Oak Ridge National Laboratory Acknowledgements: Jeff Harris, Hideo Sugama, Shin Nishimura, Andrew Ware, Steve Hirshman, Wayne Houlberg, Jim Lyon Innovative Concepts Workshop Austin, Texas February
Non-inductive plasma startup and current profile modification in Pegasus spherical torus discharges
Non-inductive plasma startup and current profile modification in Pegasus spherical torus discharges Aaron J. Redd for the Pegasus Team 2008 Innovative Confinement Concepts Workshop Reno, Nevada June 24-27,
Progress Toward High Performance Steady-State Operation in DIII D
Progress Toward High Performance Steady-State Operation in DIII D by C.M. Greenfield 1 for M. Murakami, 2 A.M. Garofalo, 3 J.R. Ferron, 1 T.C. Luce, 1 M.R. Wade, 1 E.J. Doyle, 4 T.A. Casper, 5 R.J. Jayakumar,
The Linear Theory of Tearing Modes in periodic, cyindrical plasmas. Cary Forest University of Wisconsin
The Linear Theory of Tearing Modes in periodic, cyindrical plasmas Cary Forest University of Wisconsin 1 Resistive MHD E + v B = ηj (no energy principle) Role of resistivity No frozen flux, B can tear
D.J. Schlossberg, D.J. Battaglia, M.W. Bongard, R.J. Fonck, A.J. Redd. University of Wisconsin - Madison 1500 Engineering Drive Madison, WI 53706
D.J. Schlossberg, D.J. Battaglia, M.W. Bongard, R.J. Fonck, A.J. Redd University of Wisconsin - Madison 1500 Engineering Drive Madison, WI 53706 Concept Overview Implementation on PEGASUS Results Current
The RFP: Plasma Confinement with a Reversed Twist
The RFP: Plasma Confinement with a Reversed Twist JOHN SARFF Department of Physics University of Wisconsin-Madison Invited Tutorial 1997 Meeting APS DPP Pittsburgh Nov. 19, 1997 A tutorial on the Reversed
Recent Development of LHD Experiment. O.Motojima for the LHD team National Institute for Fusion Science
Recent Development of LHD Experiment O.Motojima for the LHD team National Institute for Fusion Science 4521 1 Primary goal of LHD project 1. Transport studies in sufficiently high n E T regime relevant
MHD Stabilization Analysis in Tokamak with Helical Field
US-Japan Workshop on MHD Control, Magnetic Islands and Rotation the University of Texas, Austin, Texas, USA AT&T Executive Education & Conference Center NOVEMBER 3-5, 8 MHD Stabilization Analysis in Tokamak
MHD-Induced Alpha Particle Loss in TFTR. S.J. Zweben, D.S. Darrow, E.D. Fredrickson, G. Taylor, S. von Goeler, R.B. White
MHD-Induced Alpha Particle Loss in TFTR S.J. Zweben, D.S. Darrow, E.D. Fredrickson, G. Taylor, S. von Goeler, R.B. White Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 8543 Abstract MHD-induced
Comparison of Divertor Heat Flux Splitting by 3D Fields with Field Line Tracing Simulation in KSTAR
1 Comparison of Divertor Heat Flux Splitting by 3D Fields with Field Line Tracing Simulation in KSTAR W. Choe 1,2*, K. Kim 1,2, J.-W. Ahn 3, H.H. Lee 4, C.S. Kang 4, J.-K. Park 5, Y. In 4, J.G. Kwak 4,
Rotation and Neoclassical Ripple Transport in ITER
Rotation and Neoclassical Ripple Transport in ITER Elizabeth J. Paul 1 Matt Landreman 1 Francesca Poli 2 Don Spong 3 Håkan Smith 4 William Dorland 1 1 University of Maryland 2 Princeton Plasma Physics
Control of linear modes in cylindrical resistive MHD with a resistive wall, plasma rotation, and complex gain
Control of linear modes in cylindrical resistive MHD with a resistive wall, plasma rotation, and complex gain Dylan Brennan 1 and John Finn 2 contributions from Andrew Cole 3 1 Princeton University / PPPL
GA A26887 ADVANCES TOWARD QH-MODE VIABILITY FOR ELM-FREE OPERATION IN ITER
GA A26887 ADVANCES TOWARD QH-MODE VIABILITY FOR ELM-FREE OPERATION IN ITER by A.M. GAROFALO, K.H. BURRELL, M.J. LANCTOT, H. REIMERDES, W.M. SOLOMON and L. SCHMITZ OCTOBER 2010 DISCLAIMER This report was
Effect of ideal kink instabilities on particle redistribution
Effect of ideal kink instabilities on particle redistribution H. E. Ferrari1,2,R. Farengo1, P. L. Garcia-Martinez2, M.-C. Firpo3, A. F. Lifschitz4 1 Comisión Nacional de Energía Atómica, Centro Atomico
Influence of ECR Heating on NBI-driven Alfvén Eigenmodes in the TJ-II Stellarator
EX/P- Influence of ECR Heating on NBI-driven Alfvén Eigenmodes in the TJ-II Stellarator Á. Cappa, F. Castejón, T. Estrada, J.M. Fontdecaba, M. Liniers and E. Ascasíbar Laboratorio Nacional de Fusión CIEMAT,
Resistive wall mode stabilization by slow plasma rotation in DIII-D tokamak discharges with balanced neutral beam injection a
PHYSICS OF PLASMAS 14, 056101 2007 Resistive wall mode stabilization by slow plasma rotation in DIII-D tokamak discharges with balanced neutral beam injection a E. J. Strait, b A. M. Garofalo, c G. L.
Overview of Pilot Plant Studies
Overview of Pilot Plant Studies and contributions to FNST Jon Menard, Rich Hawryluk, Hutch Neilson, Stewart Prager, Mike Zarnstorff Princeton Plasma Physics Laboratory Fusion Nuclear Science and Technology
Heating and Current Drive by Electron Cyclotron Waves in JT-60U
EX/W- Heating and Current Drive by Electron Cyclotron Waves in JT-6U T. Suzuki ), S. Ide ), C. C. Petty ), Y. Ikeda ), K. Kajiwara ), A. Isayama ), K. Hamamatsu ), O. Naito ), M. Seki ), S. Moriyama )
Characterization of neo-classical tearing modes in high-performance I- mode plasmas with ICRF mode conversion flow drive on Alcator C-Mod
1 EX/P4-22 Characterization of neo-classical tearing modes in high-performance I- mode plasmas with ICRF mode conversion flow drive on Alcator C-Mod Y. Lin, R.S. Granetz, A.E. Hubbard, M.L. Reinke, J.E.
Princeton Plasma Physics Laboratory. Multi-mode analysis of RWM feedback with the NMA Code
Princeton Plasma Physics Laboratory Multi-mode analysis of RWM feedback with the NMA Code M. S. Chance, M.Okabayashi, M. S. Chu 12 th Workshop on MHD Stability Control: Improved MHD Control Configurations
Edge Rotational Shear Requirements for the Edge Harmonic Oscillation in DIII D Quiescent H mode Plasmas
Edge Rotational Shear Requirements for the Edge Harmonic Oscillation in DIII D Quiescent H mode Plasmas by T.M. Wilks 1 with A. Garofalo 2, K.H. Burrell 2, Xi. Chen 2, P.H. Diamond 3, Z.B. Guo 3, X. Xu
Supported by. Role of plasma edge in global stability and control*
NSTX Supported by College W&M Colorado Sch Mines Columbia U CompX General Atomics INL Johns Hopkins U LANL LLNL Lodestar MIT Nova Photonics New York U Old Dominion U ORNL PPPL PSI Princeton U Purdue U
Direct drive by cyclotron heating can explain spontaneous rotation in tokamaks
Direct drive by cyclotron heating can explain spontaneous rotation in tokamaks J. W. Van Dam and L.-J. Zheng Institute for Fusion Studies University of Texas at Austin 12th US-EU Transport Task Force Annual
D.J. Schlossberg, D.J. Battaglia, M.W. Bongard, R.J. Fonck, A.J. Redd. University of Wisconsin - Madison 1500 Engineering Drive Madison, WI 53706
D.J. Schlossberg, D.J. Battaglia, M.W. Bongard, R.J. Fonck, A.J. Redd University of Wisconsin - Madison 1500 Engineering Drive Madison, WI 53706 Non-solenoidal startup using point-source DC helicity injectors
Resistive Wall Mode Observation and Control in ITER-Relevant Plasmas
Resistive Wall Mode Observation and Control in ITER-Relevant Plasmas J. P. Levesque April 12, 2011 1 Outline Basic Resistive Wall Mode (RWM) model RWM stability, neglecting kinetic effects Sufficient for
A simple model of the resistive wall mode in tokamaks
A simple model of the resistive wall mode in tokamaks Richard Fitzpatrick Institute for Fusion Studies, Department of Physics, University of Texas at Austin, Austin TX 78712 (February 18, 2003) A simple
MAGNETIC FIELD ERRORS: RECONCILING MEASUREMENT, MODELING AND EMPIRICAL CORRECTIONS ON DIII D
MAGNETIC FIELD ERRORS: RECONCILING MEASUREMENT, MODELING AND EMPIRICAL CORRECTIONS ON DIII D by M.J. Schaffer, T.E. Evans, J.L. Luxon, G.L. Jackson, J.A. Leuer, J.T. Scoville Paper QP1.76 Presented at
EFFECT OF PLASMA FLOWS ON TURBULENT TRANSPORT AND MHD STABILITY*
EFFECT OF PLASMA FLOWS ON TURBULENT TRANSPORT AND MHD STABILITY* by K.H. BURRELL Presented at the Transport Task Force Meeting Annapolis, Maryland April 3 6, 22 *Work supported by U.S. Department of Energy
Characteristics of the H-mode H and Extrapolation to ITER
Characteristics of the H-mode H Pedestal and Extrapolation to ITER The H-mode Pedestal Study Group of the International Tokamak Physics Activity presented by T.Osborne 19th IAEA Fusion Energy Conference
HIGH PERFORMANCE EXPERIMENTS IN JT-60U REVERSED SHEAR DISCHARGES
HIGH PERFORMANCE EXPERIMENTS IN JT-U REVERSED SHEAR DISCHARGES IAEA-CN-9/EX/ T. FUJITA, Y. KAMADA, S. ISHIDA, Y. NEYATANI, T. OIKAWA, S. IDE, S. TAKEJI, Y. KOIDE, A. ISAYAMA, T. FUKUDA, T. HATAE, Y. ISHII,
ITER operation. Ben Dudson. 14 th March Department of Physics, University of York, Heslington, York YO10 5DD, UK
ITER operation Ben Dudson Department of Physics, University of York, Heslington, York YO10 5DD, UK 14 th March 2014 Ben Dudson Magnetic Confinement Fusion (1 of 18) ITER Some key statistics for ITER are:
Current Drive Experiments in the HIT-II Spherical Tokamak
Current Drive Experiments in the HIT-II Spherical Tokamak T. R. Jarboe, P. Gu, V. A. Izzo, P. E. Jewell, K. J. McCollam, B. A. Nelson, R. Raman, A. J. Redd, P. E. Sieck, and R. J. Smith, Aerospace & Energetics
Introduction to Fusion Physics
Introduction to Fusion Physics Hartmut Zohm Max-Planck-Institut für Plasmaphysik 85748 Garching DPG Advanced Physics School The Physics of ITER Bad Honnef, 22.09.2014 Energy from nuclear fusion Reduction
AC loop voltages and MHD stability in RFP plasmas
AC loop voltages and MHD stability in RFP plasmas K. J. McCollam, D. J. Holly, V. V. Mirnov, J. S. Sar, D. R. Stone UW-Madison 54rd Annual Meeting of the APS-DPP October 29th - November 2nd, 2012 Providence,
W.A. HOULBERG Oak Ridge National Lab., Oak Ridge, TN USA. M.C. ZARNSTORFF Princeton Plasma Plasma Physics Lab., Princeton, NJ USA
INTRINSICALLY STEADY STATE TOKAMAKS K.C. SHAING, A.Y. AYDEMIR, R.D. HAZELTINE Institute for Fusion Studies, The University of Texas at Austin, Austin TX 78712 USA W.A. HOULBERG Oak Ridge National Lab.,
Observation of Neo-Classical Ion Pinch in the Electric Tokamak*
1 EX/P6-29 Observation of Neo-Classical Ion Pinch in the Electric Tokamak* R. J. Taylor, T. A. Carter, J.-L. Gauvreau, P.-A. Gourdain, A. Grossman, D. J. LaFonteese, D. C. Pace, L. W. Schmitz, A. E. White,
Evolution of Bootstrap-Sustained Discharge in JT-60U
1 Evolution of Bootstrap-Sustained Discharge in JT-60U Y. Takase 1), S. Ide 2), Y. Kamada 2), H. Kubo 2), O. Mitarai 3), H. Nuga 1), Y. Sakamoto 2), T. Suzuki 2), H. Takenaga 2), and the JT-60 Team 1)
On the physics of shear flows in 3D geometry
On the physics of shear flows in 3D geometry C. Hidalgo and M.A. Pedrosa Laboratorio Nacional de Fusión, EURATOM-CIEMAT, Madrid, Spain Recent experiments have shown the importance of multi-scale (long-range)
Transport Improvement Near Low Order Rational q Surfaces in DIII D
Transport Improvement Near Low Order Rational q Surfaces in DIII D M.E. Austin 1 With K.H. Burrell 2, R.E. Waltz 2, K.W. Gentle 1, E.J. Doyle 8, P. Gohil 2, C.M. Greenfield 2, R.J. Groebner 2, W.W. Heidbrink
Energetic-Ion-Driven MHD Instab. & Transport: Simulation Methods, V&V and Predictions
Energetic-Ion-Driven MHD Instab. & Transport: Simulation Methods, V&V and Predictions 7th APTWG Intl. Conference 5-8 June 2017 Nagoya Univ., Nagoya, Japan Andreas Bierwage, Yasushi Todo 14.1MeV 10 kev
DT Fusion Ignition of LHD-Type Helical Reactor by Joule Heating Associated with Magnetic Axis Shift )
DT Fusion Ignition of LHD-Type Helical Reactor by Joule Heating Associated with Magnetic Axis Shift ) Tsuguhiro WATANABE National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan (Received
Tokamak/Helical Configurations Related to LHD and CHS-qa
9TH WORKSHOP ON MHD STABILITY CONTROL: "CONTROL OF MHD STABILITY: BACK TO THE BASICS" NOVEMBER 21-23, 2004, PRINCETON PLASMA PHYSICS LABORATORY Tokamak/Helical Configurations Related to LHD and CHS-qa
Magnetohydrodynamics (MHD) II
Magnetohydrodynamics (MHD) II Yong-Su Na National Fusion Research Center POSTECH, Korea, 8-10 May, 2006 Review I 1. What is confinement? Why is single particle motion approach required? 2. Fluid description
Flow and dynamo measurements in the HIST double pulsing CHI experiment
Innovative Confinement Concepts (ICC) & US-Japan Compact Torus (CT) Plasma Workshop August 16-19, 211, Seattle, Washington HIST Flow and dynamo measurements in the HIST double pulsing CHI experiment M.
Highlights from (3D) Modeling of Tokamak Disruptions
Highlights from (3D) Modeling of Tokamak Disruptions Presented by V.A. Izzo With major contributions from S.E. Kruger, H.R. Strauss, R. Paccagnella, MHD Control Workshop 2010 Madison, WI ..onset of rapidly
Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT
Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT S. C. Jardin 1, C. E. Kessel 1, T. K. Mau 2, R. L. Miller 2, F. Najmabadi 2, V. S. Chan 3, M. S. Chu 3, L. L. Lao 3, T. Petrie 3, P. Politzer
GA A27857 IMPACT OF PLASMA RESPONSE ON RMP ELM SUPPRESSION IN DIII-D
GA A27857 IMPACT OF PLASMA RESPONSE ON RMP ELM SUPPRESSION IN DIII-D by A. WINGEN, N.M. FERRARO, M.W. SHAFER, E.A. UNTERBERG, T.E. EVANS, D.L. HILLIS, and P.B. SNYDER JULY 2014 DISCLAIMER This report was