Fusion Development Facility (FDF) Divertor Plans and Research Options A.M. Garofalo, T. Petrie, J. Smith, M. Wade, V. Chan, R. Stambaugh (General Atomics) J. Canik (Oak Ridge National Laboratory) P. Stangeby (University of Toronto) M.V. Umansky (LLNL) Joint Workshop ReNeW Theme III and Theme IV Theme III - Plasma Material Interface Theme IV - Fusion Power March 2-6, 2009 UCLA Los Angeles, California
FDF Divertor Requirements Heat flux Dissipation of <10 MW/m 2 in steady-state (similar to ITER) Fuel and Helium Exhaust Adequate fuel exhaust for density control Adequate helium exhaust to maintain ƒ He < 10% Core performance Compatibility with good core confinement and stability in order to achieve high level mission goals Geometry Compact design for contained machine cost Plasma material interface Compatibility with high neutron flux
Baseline Design for FDF Is a Balanced Double-null Divertor Configuration Higher triangularity is achieved more naturally in balanced doublenull (DN) configuration Stability and confinement improve with increasing triangularity DN configuration reduces peak heat flux by increasing total wetted area of divertor surfaces Particle and heat flux in DN operation is predominantly to the outboard divertor (~80%) No need to install pumping hardware on difficult to engineer high-field side of plasma Requires less heat protective armor at the inner divertor targets DN choice requires shaping/control coils to accurately and robustly maintain divertor balance from relatively far away from the X-points
Peak Divertor Heat Flux < 10 MW/m 2 Estimated Using Multi-machine, Multi-parameter Scaling Estimate of exponential power flux width: Integral power flux width: λ q (Loarte 1999) Exponential heat flux width λ exp = 0.5 x λ q Estimate of peak heat flux at the divertor target (assuming zero SOL and divertor radiation): Q peak out = (P heat " P rad )F OD sin# N D 2$R D F X % exp FDF 2 MW/m 2 case (baseline) P heat =108 MW, B T =6 T, q 95 =5 " exp = Q peak out = 7.2 mm 8.3 MW/m 2 Radiated power in the core P rad =54% P heat Number of divertors N D =2 for double-null operation Fraction of power to outer divertor F OD =0.8: Poloidal tilt angle between field lines and divertor plate θ=10 Major radius at divertor strike point, R D =2.3 m Flux expansion at divertor strike point, F X =4
Large Scatter in Power Dependence of Heat Flux Profile Width from Various Empirical Scalings Chapter 4 of Progress in the ITER Physics Basis
Large Scatter in Power Dependence of Heat Flux Profile Width from Various Empirical Scalings Chapter 4 of Progress in the ITER Physics Basis Reasons for scatter include: Difficulties in diagnostic interpretation Poorly constrained data sets Insufficient number of data points There is an urgent need for improved experimental measurements, improved database, and improved coupling to theory for making reliable extrapolations for the target heat flux in FDF, or DEMO
A Number of Factors Can Alter the Peak Heat Flux From Simple Field Line Following Estimates Factor Increase/Decrease Increased Core Plasma Radiation Down 1.3x SOL and Divertor Radiation Down 4-5x Increasing Flux Expansion at the Target Plate, X Divertor Down 2x Snowflake Type Divertor Down 1.6x Super X Divertor Down at least 2x Slow Rotation of the Distorted Heat Flux Pattern Produced by the RMP Down 1.5x Coils to Suppress ELMs Less Plate Tilt Angle, perhaps 20 deg Up 2x Shorter Midplane Heat Flux Falloff Length by Factor 2 Up 2x Excursions in Control from DN to SN Up 1.6x } 2x Maximum cumulative potential increase of order a factor 6 Maximum cumulative potential reduction of order a factor 18 Considering Snowflake, X, and SX divertor options as forms of flux expansion
DIII-D Shows Radiative Dissipation and Stochastic Edge Viable Approaches to Reducing Peak Heat Flux Divertor radiation preferentially enhanced using puff and pump technique P rad /P NBI ~ 60% with Z eff ~2.0 Argon and D2 injection β N =2.6, H 89 =2.0, G=0.4 maintained Experiments thus far focused on puff and pump studies, not fusion performance Major Radius R (m) Resonant Magnetic Perturbation splits strike points Angular rotation of the RMP will result in a time averaged broadening of the OSP footprint
A Closed Geometry to Best Take Advantage of Radiative Dissipation with Tilted Divertor Closure to impede mobility of neutrals away from divertor and facilitate pumping UEDGE modeling showed bad synergy between tilted plate and impurity radiation in open divertor Up/down symmetric design, allowing pumping from both private flux region and outboard side Optimal design may not be exactly up/down symmetric because of effect of ExB drifts Tilted divertor plate Angle between magnetic field line and divertor tile surface ~1º Divertor must present a perfectly toroidally flat surface to field lines FDF equilibrium
Precision Alignment Enables Full Use of Divertor Plate Tilting and Flux Expansion Divertor can be built in a factory setting as a single precision toroidal ring with precision placement of tiles on it DIII-D lower divertor Slight fish-scaling of the tiles may be done to hide edges Problem of heat flux handling becomes axisymmetric Because of precision alignment, infrared thermography pictures of the divertor in DIII-D show near-total axisymmetry Outer strike point Nose tiles of pump baffle (on floor near pump baffle)
Precision Alignment Enables Full Use of Divertor Plate Tilting and Flux Expansion Divertor can be built in a factory setting as a single precision toroidal ring with precision placement of tiles on it DIII-D lower divertor heat flux before and after Improved tile alignment Slight fish-scaling of the tiles may be done to hide edges Problem of heat flux handling becomes axisymmetric Because of precision alignment, infrared thermography pictures of the divertor in DIII-D show near-total axisymmetry
SOLPS Allows Modeling of FDF s Closed Divertor Design SOLPS: Scrape Off Layer Plasma Simulation 2D plasma fluid code (B2.5) Monte Carlo neutrals code (Eirene) Carbon sputtering at all surfaces, constant chemical yield of 1% D 2 puff at midplane Pumping at ~50 m 3 /s Transport model Classical parallel to B (+flux limiters) ITER heat and particle diffusion coefficients in the SOL, lower in the pedestal Core boundary conditions Total Heating Power = 108.14 MW Brehmsstrahlung loss = 13.19 MW Core radiation = 45.42 MW Power into SOL = 49.53 MW
SOLPS Modeling Shows Strong Reduction of Peak Heat Flux with Increasing Gas Puff Low gas-puff case is conductionlimited case to compare heat flux width to empirical scalings Gas-puff raised just until point where there is some parallel T e gradient λ exp,omp ~ 3 mm T e,osp ~ 150 ev, P rad =16 MW Medium gas puff case λ exp,omp ~ 3 mm T e,osp ~ 12 ev, P rad =33 MW High gas puff case (nearly detached) λ exp,omp ~ 10 mm T e,osp ~ 2 ev, P rad =39 MW
SOLPS Results Show SOL and Divertor Radiation Fairly Well Dispersed Away From Divertor Plate P rad (W/m 3 ) Z (m) High gas puff case Impurity and hydrogenic radiation Small contribution to power onto target R (m)
DIII-D Will Carry Out Tests of Various Advanced Divertor Approaches for Increasing Flux Expansion X-divertor coils being evaluated for DIII-D DIII-D Snowflake Divertor experiment proposed for FY09 M.V. Umansky, et al., LLNL I/I max = 92% DIII-D plasma with 2nd x-point set 5 cm below primary x-point 120% 75% 53% Kotschenreuther et al., Phys. Plasmas, 2007 DIII-D Super-X Divertor experiment proposed for FY10 P. Valanju et al., IFS, Univ. of Texas SXD obtained for FDF with just one extra PF coil
FDF Is Well Positioned to Address DEMO Divertor Design Issues Results from both simple and 2-D plasma modeling analysis predict peak heat flux < 10 MW/m 2 at outer divertor targets ~50% core radiation fraction -> Psol/Rdiv ~21.7 (FDF) vs. ~18 (ITER) Well aligned tiles -> B-target angle ~1º (FDF) vs. ~2.5º (ITER) Several approaches can be used in parallel to further reduce peak heat flux Flux expansion (X divertor, SX divertor, Snowflake divertor) SOL and divertor radiative dissipation Rotating stochastic edge Expected progress: Refinements to methodologies, including addition of particle drifts to SOLPS analyses New, more reliable empirical scalings likely to result from Joint (C-Mod, NSTX, DIII-D) Joule Milestone for 2010 on SOL thermal transport physics Experimental tests of Snowflake and SXD concepts Refinements to divertor design, e.g. up/down asymmetry to account for effect of ExB drifts
Plan for Achievement of Excellent Tile Alignment in FDF Is Key to Manageable Peak Power Angle between magnetic field and target surface at outer strike point in ITER is ~ 2.5 deg, while it is only ~ 1 deg in FDF FDF angle between field and surface ITER old (2001) geometry ITER latest geometry 4.0 3.0 angle [degrees] 2.0 1.0 0.0-0.100 0.000 0.100 0.200 0.300 S-S_OSP [m]
Why Peak Power Deposition in FDF Is Calculated to Be ~ ITER, Even Though FDF Is Higher Power Density If ~50% core radiation fraction can be achieved and targets can be aligned precisely, heat loading problem in FDF should be much easier than ITER Use Psol/Rdiv (maybe not a great figure of merit, but more relevant to divertor loading than P/R) Psol/Rdiv ~ 18 for ITER Psol/Rdiv ~ 21.7 for FDF (i.e. not very different) In addition: DN (FDF) vs. SN (ITER) divertor configuration FDF divertor plates are twice as tilted => Reasonable to expect lower heat fluxes in FDF than in ITER