Core Design Derek Sutherland, Cale Kasten Choongki Sung, Tim Palmer Paul Bonoli, Dennis Whyte 22.63 - May 17, 2012
Primary Design Goals Qp ~ 25 and Qe > 3, with thermal output of ~ 500 MW. Develop a robust, steady-state scenario built from I-Mode profile scalings. Achieve high current drive efficiency at mid-radius. Achieve non-inductive scenario with modest bootstrap fraction for substantial external plasma current profile control.
Primary Design Solutions High-field allows for high power densities at normalized beta of ~ 2.5. Compact, robust fusion reactor with Qp ~ 25. High-field side launch of LH waves. Highest possible current drive efficiency at mid-radius. Low normalized beta and high-field CD launch. 20 % external control of plasma current profiles with 4% of the total fusion power. Minimize recirculating power with superconducting coils and efficient current drive. Qe > 5! with thermal output of ~511 MW.
0-D design considerations provide starting point for optimization N apple 3
0-D optimization provides range of valid solutions in R-ε space for non-inductive scenario f bs <T e > (kev) Blanket power density [MW/m 3 ]
High magnetic fields allow for compact, high power density fusion reactors. * 13 Tesla peak on coil Blanket power density [MW/m 3 ] 6
High magnetic fields allow for compact, high power density fusion reactors. * 18 Tesla peak on coil Blanket power density [MW/m 3 ] 7
High magnetic fields allow for compact, high power density fusion reactors. * 22 Tesla peak on coil Blanket power density [MW/m 3 ] 8
Plasma operating contour indicates that operating point is accessible and stable. Contours of required Pext [MW] Contours of fusion power [MW] Contours of Qp H89 ~ 2.4 Operating point 9
I-mode is chosen for its desirable heat and particle confinement properties. L-mode like particle confinement H-mode like energy confinement. No density pedestal. Does not require ELMs to regulate core impurities or pedestal. 10
n e scaling is based on C-Mod I-mode profiles in the absence of core sawteeth. 0 0.5 1 1.5 2 2.5 adapted from Whyte APS-DPP 2011 design C-Mod ne (10 20 m -3 ) Density vs. Normalized Radius scaled C-Mod data. functional fit for code input 0 0.2 0.4 0.6 0.8 1 ρ (r/a) set a constant gradient in n/n 0 n 0 scaled until 90% Greenwald fraction is achieved 10
T e scaling is based on recent C-Mod heat flux scalings with additional current scaling
Since scaling depends on α-heating power, profiles must be self-consistently iterated. Scaling Method: 1. Set P heat = P ext = 20 MW 2. Scale profiles (P heat /S), compute P f 3. Compute P α = P f /5, recompute P heat = P ext + P α 4. Iterate to constant P f 5. (choose q 95 such that P f ~ 500 MW) assumed T i = T e Boundary condition is T(a)=0.2 kev. scaled C-Mod data. functional fit for code input Te (kev) 0 5 10 15 20 25 30 Temperature vs. Normalized Radius 0 0.2 0.4 0.6 0.8 1 ρ (r/a) 12
Efficient lower hybrid current drive at mid-radius requires high-field-side launch. LHCD efficiency proportional to 1/N 2 Accessibility condition Low Field Side Launch p n N / B Advantageous to launch in region of high B with high launch frequency. High field side launch of 8.0 GHz LH waves. [1] Y. A. Podpaly et al, Fusion Engineering and Design, 86, 810 (2011) f RF High Field Side Launch 12
ACCOME code used to self-consistently calculate an equilibrium, damping locations, and current profiles. 2D, self-consistent, free-boundary, magnetic equilibrium solver. 90% of power absorbed Uses given n,t profiles Varied launched N, vertical launcher position, and launch frequency to optimize. Robust mid-axis damping while launching near region of high flux expansion. Coil positions verified. 15
High-field side LHCD provides near ideal current drive efficiency at mid-radius Robust q profile. Accessible N Ray N N ~ 1.5 is damping at 10 kev. Cannot push N lower due to accessibility and fast-wave conversion concerns. 10 kev volume averaged reactor optimal for efficient LHCD at midradius. Favors smaller reactors. 16
ACCOME predicts a successful reactor operating scenario while maintaining significant plasma control. Parameter Result Fusion Power 511 MW LHCD Coupled Power 20 MW Qp 25 BT Ip * ICD 9.2 T 7.66 MA 1.26 MA fbs 83.6% ηcd 0.37 x 10 20 AW -1 m -2 q95 ~6 * Subtracted artificial neutral beam current. Future work will include FWCD. 17
Power balance calculation indicates high plant efficiency. Q e = P net P ext = th(p fusion + P heat + P dissipated ) P ext,e Q e = th((1 + M n )P fusion + P heat + P dissipated ) + P LH + P pump P coils e Parameter Result Qe 5.12 Pth Pe 640 MW* 270 MW Plant efficiency 42% *additional 130 MW provided by Li reactions in the blanket 18
Integrated, high-field side, LHCD design minimizes waveguide losses and compensates for potential failures. Two, toroidally continuous strips of passive-active waveguides. Total installed unconditioned power 23 MW. Attenuation losses proportional to resistivity of waveguide. Power Throughput Location Wall plug Klystrons Cold waveguide Hot waveguide RF launcher Transmitted Power 55.6 MW 27.8 MW 24.0 MW 22.4 MW 20.0 MW 19
Other Design Limitations and Solutions Maximum βn limited by qmin. Intrinsic plasma rotation suppresses RWMs. Rotational speed of 650 km s -1 MA ~ 0.07 Waveguide losses in hot waveguides. Minimize length of hot waveguides. LHCD frequency must be > 7.0 GHz to avoid damping on alpha particles at mid-radius. Launch 8.0 GHz LH waves. Advantageous for LHCD efficiency as well. ELMs in reactor relevant I-Mode. Pedestal pressure stable against ELMs. 20
Future Research Priorities LHCD: Build and test waveguides and launchers in DT neutron fields to address material degradation concerns. Determine if on-axis fast-wave current drive is required for this design. Demonstrate reliable, steady-state 8.0 GHz RF sources. Demonstrate I-Mode with non-inductive profiles.
Magnet System Design J. Goh, F. Mangiarotti, S. Arsenyev, P. Le, B. Nield, D. Whyte, L. Bromberg, J. Minervini, M. Takayasu, and the rest of the ARC Reactor EDA Working Group May 17, 2012
Objectives Get B T = 9.2T on axis, within space, structural, and current density limits Be able to disassemble the TF coils for vacuum vessel removal Get enough magnetic flux swing for start up Minimize cool down and warm up time
Permanent SC Coils Conceptual design: we use two kinds of coils Large currents, large background B Use subcooled YBCO Stresses supported by SS 316 Replaceable Copper Coils For plasma shaping and control Small currents for low power losses (~1MW) FLiBe cooled, close to vacuum vessel
Subcooled YBCO can carry large currents at high magnetic fields At 4.2K: 25T 25T Our extrapolations yield J E ~400A/mm 2 @ 25T, 20K. We use ~320A/mm 2 = our design Adapted from: Drew W. Hazelton, Applications Using SuperPower 2G HTS Conductor, 2011 CEC/ICMC Conference, Spokane, WA
Toroidal Field Coils Design trade-offs Shape of magnet: D shape: fewer joints, lower resistance, less structure but is more complicated C Mod window frame style : more joints, higher resistance, more structure but is simpler Style of joint: Comb : Requires tapes to be parallel, allows better joint resistance. But: More complicated, and joints transmit strain. Edge : Simpler. Tapes can be perpendicular, and some joint movement is tolerated (no strain). But: Higher resistance.
Both TF shapes are compared for joint and structural performance D Shape Window Shape Central Column is Similar in Both Shapes
Comb Joint Geometry Minimizes Joint Resistance Joint Area ~40x larger than lap joint Preliminary experimental resistance data is available Disconnected Connected
D-Shape is Modified to Include Straightened Joint Expansion
1.5 kw of Joint Cooling Per Coil 50W/m 2 on tape, extrapolating from measured data ~1 m/s liquid H 2 flow in 1mm OD channels in comb spikes Peak Temperature: 23K Total cooling power: ~1.5 kw per TF coil Temperature Distribution in Comb Joint Unit Cell HTS Tapes Steel Structure Cooling Channel
Twist-and-Lift To Demount Top Leg
Demountability with High Field Requires Structural Innovation Large rings support outward forces normally taken in tension Fiberglass/epoxy plug takes compressive forces on inside
Analysis Shows Acceptable Safety Factor in SS316 w/ 9.2T On-Axis SF ~ 1.5 in worst areas 1 GPa Bottom Top 0.5 GPa 0.1 GPa
Window-Shape: SS316 Structure holds Lorentz forces in Coils Structure: Top Plate Top legs of TF Tube like structure on the side Central structure: equal to D shape
Acceptable stress levels on Window- Shape structure with 9.2T On-Axis 1 GPa 0.5 GPa Analyzed stresses on the green section: 10 o cut of structure and coils Worst area: 700 MPa 0 GPa
Perpendicular edge joints allow strainless connection But joint heating is higher than comb style: Theoretical prediction: 20 kw/coil Extrapolation from experiment: 200 kw/coil We use 40 kw/coil guess
Power considerations vs. ease of design: power plant vs. FNSF D Shape Window Shape Joint dissipation 30 kw 720 kw @ LH 2 Heat radiated from FLiBe Wall plug Electric Power 160 kw (@LN 2 ) 700 W (@LH 2 ) 4.4 MW 160 kw (@LN 2 ) 700 W (@LH 2 ) 52 MW Power Plant FNSF
Efficient use of Steel: cooling and warming times are few days Cooling: amount of LN 2 Cooling: amount of LH 2 D Shape 20 trucks (600m 3 ) 6 trucks (180m 3 ) Window Shape 95 trucks (2900m 3 ) 30 trucks (900m 3 ) Warming up time: ~3 days with forced flow of dry air Cooling down time: If we dunk all that liquid at once, ~8 hours. Probably we ll cool down slower, at the same pace we warm up: ~3 days
The Central Solenoid is used to start the reactor & for shaping Conductor: YBCO in LH 2 Structure: Stainless Steel 316 Graded: 2 Inner layers: 15MA turn 2 Outer layers: 53MA turn
We reach 14 T and 34 Wb with stresses below 350 MPa Magnetic Field [T] von Mises stress [Pa] Z [m] Color scale: B.field 0-14T stress 0-350MPa R [m] R [m]
Summary We have shown that high field, demountable superconducting coils are plausible Can be made in either Window or D Shape configuration D Shape offers better joint performance and structural efficiency But window offers simplicity Research required: Properties of YBCO at ~25T, LH 2 Large scale test of YBCO joints Build ARC
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Blanket and Vessel Design Justin Ball Jen Sierchio Brandon Sorbom 22.63 May 17, 2012 Dennis Whyte, Zach Hartwig, Geoff Olynyk, Harold Barnard, Christian Haakonsen, and Pete Stahl
Blanket Group Design Challenges Protecting superconducting coils from fast neutron flux is difficult for a small design Minimize inboard blanket thickness to maximize toroidal field on axis Achieve Tritium Breeding Ratio (TBR) > 1.1 Manage material damage, temperature, and activation to vacuum vessel and blanket
Blanket Group Design Innovations Replaceable, modular integrated vacuum vessel Liquid immersion blanket of enriched FLiBe Solves problems of material and disruption damage to blanket Double-walled, FLiBe cooled vacuum vessel Critical to obtaining high TBR and eliminates He pumping requirements (H. Barnard)
Reactor Cross-Section Model used in MCNP to Assess TF Lifetime, TBR, and Damage Legend Green ZrH 2 Brown Vacuum (Insulating Gap) Dark Grey Inconel 718 Red Beryllium* Yellow Tungsten Light Blue 90% 6 Li Enriched FLiBe Dark Blue YBCO + Steel Support Pink Plasma Midplane Tungsten Neutron Shield Neutron Shield TF Coil Plasma Cooling Channel *not structural material Liquid FLiBe Blanket
An Early Simulation Set Shows the Effect of Exchanging Blanket and ZrH 2 Shield *Total inboard midplane thickness constant at 80 cm 70 TF Coil Lifetime (years) 60 50 40 30 20 10 ZrH 2 Shield 0 0 20 40 60 ZrH 2 Shield Thickness (cm)
Using Conservative Estimates, We Achieve a TF Lifetime of 31 Full Power Years (FPY) Total fast neutron fluence of 3 10 neutrons/cm 2 to YBCO tape was used as the maximum experimentally achieved limiting factor [1], although actual limit is almost definitely higher Assumed 511 MW fusion yield to calculate neutron flux Added midplane shield to optimize TF lifetime [1] L. Bromberg et al., Options for the use of high temperature superconductor in tokamak fusion reactor designs (2001)
Use 90% Enriched 6 Li FLiBe with 2cm Be Multiplier to Achieve TBR of 1.14 Goal of TBR > 1.1 to account for uncertainties in neutronics codes/cross sections, final structure FLiBe-cooled channel critical to achieving high TBR TBR very sensitive to vacuum vessel build Inner VV Thickness (cm) Blanket TBR Channel TBR Total TBR 0.5 0.931 0.263 1.194 Inner VV 1 0.890 0.268 1.158 1.5 0.864 0.276 1.140 2 0.822 0.280 1.102
With a TBR of 1.14, ARC Would Produce ~3.3 kg of T extra per Year Assumptions: TBR = 1.14 Plant Availability = 85% Fusion Power = 511 MW Using a DOE estimate [1] of $100,000/g T, ARC would produce roughly $330M worth of tritium every year. Kg T/Year Produced Avg. g T/Day Produced 27.72 75.93 Kg T/Year Used Avg. g T/Day Used 24.36 66.74 Kg T/Year Extra Avg. g T/Day Extra 3.35 9.19 [1] Modernization of Tritium Requirements Systems, DOE/IG-0632, December 2003
DPA/He Production in 1 FPY is Manageable for Replaceable Components Material Layer Alphas (appm) Again, no material damage to liquid FLiBe blanket Difference of +/- ~10% for outboard/inboard Displacements per Atom Tungsten FW 4 14 Inner VV 320 43 Outer VV 180 27 Be Multiplier 3100 15 FLiBe Blanket N/A! N/A! Tungsten Shield 0.5 4 Blanket Tank 0.1 0.02 ZrH 2 Shield 0.003 0.008
DPA/He Production is Comparable to those in IFMIF and Large Fusion Reactors He Production vs. DPA for inner VV is limiting factor Puts us in between IFMIF and a large fusion reactor Running for weeks or months allows testing of components (From S. Zinkle, assumes 2 years runtime)
COMSOL Turbulent Flow Model Determined Blanket Temperature Limits Temperature (K) and Flow Field K FLiBe Properties Low electrical conductivity Low toxicity Twice density of water Similar C p to water Similar viscosity to water Melting point : 732 K Average Outlet Temperature of 886 K Surface Heat Flux from Plasma Peak Temperature of Blanket Inlet Temperature of 800 K R Inlet Velocity of 0.1 m/s
Channel length is the poloidal circumference of 10m COMSOL Turbulent Flow Model Determined VV Temperature Limits Fixed by blanket at 1000 K Outer VV 1 cm Beryllium Multiplier 2 cm FLiBe Channel 2 cm Inner VV 1.5 cm Volumetric Neutron Heating (MW/m 3 ) 5.4 10.8 22.5 9.8 Tungsten 1 mm Heat flux of 0.17 MW/m 2 from plasma Inlet Temperature of 800 K Inlet Velocity of 4 m/s
COMSOL Turbulent Flow Model Determined VV Temperature Limits Outer VV Beryllium Multiplier FLiBe Channel Inner VV Peak temperature of VV Unirradiated Inconel 718 [1] Outlet Inlet [1] Special Metals Corp., Pub. No. SMC-045, Sept. 2007
f f I I V halo t halo j 0.6I VV Unirradiated VV Survives Worst Case, Unmitigated Disruption jhalo 1 cos B B 0.4I p V p 2.5cm Von Mises Stresses (MPa) MPa Peak stress in VV Yield Strength of unirradiated Inconel 718 at 1000K 1000 MPa Von Mises stresses scale linearly with plasma current Damage is localized to replaceable components
Integrated Design Indicates Tradeoffs in VV Mechanical Factor of Safety* 1.8 1.7 1.6 1.5 1.4 1.3 1.2 1.1 1 0.9 0.5 1 1.5 2 Inner VV Thickness (cm) 1.2 1.18 1.16 1.14 1.12 1.1 1.08 1.06 1.04 1.02 1 Tritium Breeding Ratio * for a worst case, unmitigated disruption
Critical Research Needed Experimental data on critical neutron fluence for YBCO superconducting tape TF lifetime extremely sensitive to both maximum fluence and energy cutoff 3 10 neutrons/cm 2 over 0.1 MeV is not an absolute limit, just the extent of current experiments FLiBe tritium recovery system Turn-around time estimates Experimentally verified system Thermal hydraulics experiments with FLiBe in magnetic fields
ARC is a Dual-Purpose Fusion Facility: Both FNSF and Fusion Reactor Problem: Experimental data on critical DPA and He retention for materials Wider range of materials tested to higher DPA Material stress tested in reactor-like setting True validation of MCNP estimates Again, vacuum vessel is replaceable, allowing for easy testing of different designs Can also test core group research needs ARC Reactor is its own FNSF
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Extra Slides
Radial Build to Optimize TBR, TF Lifetime with Minimum Inboard Thickness Component Radial Distance (cm) Material Distance from Axis of Symmetry (cm) Core 110 Plasma 220 First Wall 0.1 Tungsten 219.9 VV Structure 1 1.5 Inconel 718 218.4 VV Cooling 2 90% Li-6 Enriched FLiBe 216.4 Multiplier 2 Beryllium 214.4 VV Structure 2 1 Inconel 718 213.4 Midplane Reflector 15 Tungsten 198.4 Blanket 15 90% Li-6 Enriched FLiBe 183.4 Blanket Tank 5 Inconel 718 178.4 Thermal Insulation 3 Vacuum 175.4 Neutron Shield 45 ZrH2 130.4
Alternate Multiplier Configuration Double Be multipliers on either side of Coolant Channel (1cm each) VV1 Thickness (cm) Blanket TBR Coolant TBR Total TBR 0.5 0.934 0.279 1.213 1 0.894 0.280 1.174 1.5 0.856 0.285 1.140 2 0.827 0.282 1.109
Be Thickness MCNP Results Thickness of Be (cm) Blanket TBR Coolant TBR Total TBR 0.5 0.917 0.232 1.149 1 0.907 0.262 1.170 2 0.877 0.316 1.193 3 0.852 0.368 1.220 4 0.807 0.422 1.230 5 0.759 0.460 1.219
Different Midplane Shield Materials 15 cm First Shield Material Effect on TF Lifetime 30.00 25.00 TF Lifetime (years) 20.00 15.00 10.00 Tungsten Tungsten Carbide Beryllium Graphite ZrH2 5.00 0.00
Evolution of MCNP Model
Start-up Tritium g T/day Tritium Burn-Up Fraction Days of Fuel Required Total Start-up Fuel (g) Cost of Start-Up Fuel ($) 78.52 0.01 0.04 327 32,700,000 Time to Collect Enough T to Start up another ARC reactor (days) 30 g T/day Tritium Burn-Up Fraction Days of Fuel Required Total Start-up Fuel (g) Cost of Start-Up Fuel ($) 78.52 0.01 1.00 7,852 785,200,000 Time to Collect Enough T to Start up another ARC reactor (days) 726
Tritium Recovery System Full analysis beyond the scope of this conceptual design Through a literature search, found recent 1 Japanese studies on T extraction from FLiBe using counter-current extraction tower Basic concept: Saturate FLiBe with Be to maintain TF concentration in FLiBe Pass saturated FLiBe down through series of filters with He pumped up in opposite direction TF diffuses in He, and T2 is pumped out with He and separated According to study, achieves T recovery > 99.9% 1.) S. Fukada, A design for recovery of tritium from Flibe loop in FFHR-2 (2007)
TBR Uncertainty in Cross Sections for MCNP Calculation UCLA study found 2-6% uncertainty in TBR for various materials based on uncertainties in nuclear databases 1 Closest material combination to ours (FLiBe/He/FS/Be) had TBR predicted overestimate of ~4.3% Total uncertainty subtracted from our TBR still gives a TBR of 1.07 1.) Uncertainties in Prediction of Tritium Breeding in Candidate Blanket Designs Due to Present Uncertainties in Nuclear Data Base, M.Z. Youssef et al, (1986)
MCNP Validations Simple fluence validation changed all cells to vacuum and made sure that all source neutrons accounted for DPA validated using NJOY processed damage cross sections and formula/code in Hogenbirk et al. 2008 TBR validated using simple toroidal model and comparing to UW results He validated: To be completed later
Cost Estimates for Vacuum Vessel Important because vacuum vessel will be replaced every year for general operation or sooner for testing purposes Total cost of raw materials (Inconel 718, tungsten, and beryllium) = $11.5 million C-mod is similar to ARC and can be used to estimate the cost C-Mod vacuum vessel = $600K (1987) which would be $1.2 million today Scaled to our size, C-Mod ~ $25 million Our vessel is more complicated, so $25 million < cost < $50 million
Cost Estimates for Blanket Tank and Shield Shield is only raw material cost + a little bit: $7 million < cost < $10 million Blanket tank requires some machining: Raw materials: $6 million Total: $10 million < cost < $25 million
Asymmetric Disruption - Model f halo = j halo B = I halo 1+ cos 2 R f V = j B V = j B V ê R Assumes a full, worst case, unmitigated disruption: I halo = 0.4I p I = 0.6I p ( ) ( ) t VV ĵp B 0R 0 ê R j halo B j j B V j B V B V B j B 0 = 9.2T B V = 0.8T t VV = 2.5cm j halo B j halo
Turbulent Model VV Channel Flow Blanket Flow V = 4 m s T =1000K D H = 2 t Channel = 0.04m 4605 = 5.94 10 5 kg [ K] T m s e = 0.006 ( ) kg = 2413-( 0.488 K 1 )*T m 3 Re = D H V =12833V = 51333 kg m s =1925 kg m 3 V = 0.1 m s T = 900K D H = 0.3m = 0.01 kg m s =1974 kg m 3 Re = 59214V = 5921 Turbulent Turbulent
Vacuum Vessel Temperature k INC 718 = 21 W m K x VV =1.5cm q VV = 0.3 P heating A VV = 0.17 MW m 2 q VV = k INC 718 T VV = k INC 718 T VV =121K T VV x VV T flibe + T VV 1121K < T melt =1533K