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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 plasmacurrent profile control.
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 profileswith 4% of the total fusion power.
Minimize recirculating power with superconducting coils and efficient current drive.
Qe > 5! with thermal output of ~511 MW.
Primary Design Solutions
0-D design considerations provide starting point for optimization
�N 3
0-D optimization provides range of valid solutions in R-ε space for non-inductive scenario
<Te> (keV)
fbs
Blanket power density[MW/m3]
6
* 13 Tesla peak on coil
Blanket power density[MW/m3]
High magnetic fields allow for compact, high power density fusion reactors.
7
* 18 Tesla peak on coil
Blanket power density[MW/m3]
High magnetic fields allow for compact, high power density fusion reactors.
8
* 22 Tesla peak on coil
Blanket power density[MW/m3]
High magnetic fields allow for compact, high power density fusion reactors.
9
Plasma operating contour indicates that operating point is accessible and stable.
Contours of required Pext [MW]Contours of fusion power [MW]Contours of QpH89 ~ 2.4
Operating point
10
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 toregulate core impurities or pedestal.
10
ne scaling is based on C-Mod I-mode profiles in the absence of core sawteeth.
C-Mod
design
set a constant gradient in n/n0
n0 scaled until 90% Greenwald fraction is achieved
adapted from Whyte APS-DPP 2011
scaled C-Mod data.functional fit for code input
Density vs. Normalized Radius
ρ (r/a)
n e (
1020
m-3)
0
0.5
1
1.
5
2
2
.50 0.2 0.4 0.6 0.8 1
Te scaling is based on recent C-Mod heat flux scalings with additional current scaling
12
assumed Ti = Te
Boundary condition is T(a)=0.2 keV.
scaled C-Mod data.functional fit for code input
Scaling Method:1. Set Pheat = Pext = 20 MW2. Scale profiles (Pheat/S), compute Pf
3. Compute Pα = Pf/5, recompute Pheat = Pext + Pα 4. Iterate to constant Pf
5. (choose q95 such that Pf ~ 500 MW)
Since scaling depends on α-heating power, profiles must be self-consistently iterated.
0 0.2 0.4 0.6 0.8 1ρ (r/a)
Te
(keV
)
0
5
10
1
5
20
25
30
Temperature vs. Normalized Radius
N|| /pn
B
12
Efficient lower hybrid current drive at mid-radius requires high-field-side launch.
LHCD efficiency proportional to 1/N||2
Accessibility condition
LowField SideLaunch
HighField SideLaunch
Advantageous to launch in region ofhigh B with high launch frequency . fRF
High field side launch of 8.0 GHz LH waves.
[1] Y. A. Podpaly et al, Fusion Engineering and Design, 86, 810 (2011)
15
ACCOME code used to self-consistently calculate an equilibrium, damping locations, and current profiles.
2D, self-consistent, free-boundary, magnetic equilibrium solver.
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.
90% of power absorbed
16
N|| ~ 1.5 is damping at 10 keV.
Cannot push N|| lower due to accessibility and fast-wave conversion concerns.
Accessible N||Ray N||
10 keV volume averaged reactoroptimal for efficient LHCD at mid-radius.
High-field side LHCD provides near ideal current drive efficiency at mid-radius Robust q profile.
Favors smaller reactors.
17
ACCOME predicts a successful reactor operating scenario while maintaining significant plasma control.
Parameter Result
Fusion Power 511 MWLHCD Coupled Power 20 MWQp 25
BT 9.2 T
Ip* 7.66 MA
ICD 1.26 MA
fBS 83.6%
ηCD 0.37 x 1020 AW-1m-2
q95 ~6
* Subtracted artificial neutral beam current. Future work will include FWCD.
Qe =⌘th((1 +Mn)Pfusion + Pheat + Pdissipated)
Pcoils
⌘e
+ PLH + Ppump
Qe
=Pnet
Pext
=⌘th
(Pfusion
+ Pheat
+ Pdissipated
)
Pext,e
18
Power balance calculation indicates high plant efficiency.
Parameter Result
Qe 5.12
Pth 640 MW*
Pe 270 MW
Plant efficiency 42%
*additional 130 MW provided by Li reactions in the blanket
19
Integrated, high-field side, LHCD design minimizes waveguide losses and compensates for potential failures.
Attenuation losses proportional to resistivity of waveguide.
Two, toroidally continuous strips of passive-active waveguides.
Total installed “unconditioned” power 23 MW.
Location Transmitted Power
Wall plug 55.6 MW
Klystrons 27.8 MW
Cold waveguide 24.0 MW
Hot waveguide 22.4 MW
RF launcher 20.0 MW
Power Throughput
• 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
Other Design Limitations and Solutions
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 DesignJ. 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 GroupMay 17, 2012
Objectives
• Get�BT =�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�startͲup
• Minimize�cool�down and�warm�up time
Conceptual design: we use two kinds of coils
• Permanent�SC�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)
– FLiBeͲcooled,�close�to�vacuum�vessel
Subcooled YBCO can carry large currents at high magnetic fields
Adapted�from:�Drew�W.�Hazelton,�Applications�Using�SuperPower 2G�HTS�Conductor,�2011�CEC/ICMC�Conference,�Spokane,�WA
25T
=�our�design
Our�extrapolations�yield�JE ~400A/mm2 @�25T,�20K.We�use�~320A/mm2
25T
At�4.2K:
Toroidal Field Coils Design trade-offs
• Shape�of�magnet:– DͲshape: fewer�joints,�lower�resistance,�less�structure�ͲͲ but�is�
more�complicated
– CͲMod�‘window�frame�style’:more�joints,�higher�resistance,�more�structure�ͲͲ 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
DͲShape WindowͲShape
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
Temperature�Distribution�in�Comb�Joint�Unit�Cell
• 50W/m2 on�tape,�extrapolating�from�measured�data• ~1�m/s liquid�H2 flow�in�1mm�OD�channels�in�comb�spikes• Peak�Temperature:�23K• Total�cooling�power:�~1.5�kW�per�TF�coil
Cooling�Channel
Steel�Structure
HTS�Tapes
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
BottomTop
1�GPa
0.1�GPa
0.5�GPa
Window-Shape: SS316 Structure holds Lorentz forces in Coils
Structure:�Top�Plate
Top�legs�of�TF
TubeͲlike�structure�on�the�side
Central�structure:�equal�to�DͲshape
Acceptable stress levels on Window-Shape structure with 9.2T On-Axis
1�GPa
0�GPa
0.5�GPa
Analyzed�stresses�on�the�green section:�10o cut�of�structure�and�coils
Worst�area:�700�MPa
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�@�LH2
30�kW 720�kW
Heat�radiated�from�FLiBe
160�kW�(@LN2)
700�W�(@LH2)
160�kW�(@LN2)
700�W�(@LH2)
Wall�plug�Electric�Power
4.4�MW 52�MW
Power�Plant FNSF
Efficient use of Steel: cooling and warming times are few days
D�Shape Window�Shape
Cooling:�amount�of�LN2
20�trucks�(600m3)
95�trucks�(2900m3)
Cooling:�amount�of�LH2
6�trucks����(180m3)
30�trucks�(900m3)
• 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�LH2
Structure:�Stainless�Steel�316
Graded:
– 2�Inner�layers:�15MAͲturn
– 2�Outer�layers:�53MAͲturn
We reach 14 T and 34 Wbwith stresses below 350 MPa
Magnetic Field [T] von Mises stress [Pa]
Color scale:B.field 0-14Tstress 0-350MPa
R [m]
Z [m]
R [m]
Summary
• We�have�shown�that�highͲfield,�demountable�superconducting�coils�are�plausible– Can�be�made�in�either�Window�or�DͲShape�configuration
– DͲShape�offers�better�joint�performance�and�structural�efficiency
– But�window�offers�simplicity
• Research�required:Properties�of�YBCO�at�~25T,�LH2
Large�scale�test�of�YBCO�joints
Build�ARC
Questions?
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 – ZrH2 Brown – Vacuum (Insulating Gap) Dark Grey – Inconel 718 Red– Beryllium* Yellow – Tungsten Light Blue – 90% 6Li Enriched FLiBe Dark Blue – YBCO + Steel Support Pink – Plasma
*not structural material
Liquid FLiBe Blanket
Neu
tron
Shi
eld
Plasma
Midplane Tungsten Neutron Shield
Cooling Channel
TF Coil
An Early Simulation Set Shows the Effect of Exchanging Blanket and ZrH2 Shield
0
10
20
30
40
50
60
70
0 20 40 60
TF C
oil L
ifetim
e (y
ears
)
ZrH2 Shield Thickness (cm)
*Total inboard midplane thickness constant at 80 cm
ZrH2
Shield
Using Conservative Estimates, We Achieve a TF Lifetime of 31 Full Power Years (FPY)
• Total fast neutron fluence of 3 × 10 neutrons/cm2 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 6Li 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
1
0.890
0.268
1.158
1.5
0.864
0.276
1.140
2
0.822
0.280
1.102
Inner VV
With a TBR of 1.14, ARC Would Produce ~3.3 kg of T extra per 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
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.
DPA/He Production in 1 FPY is Manageable for Replaceable Components
• Again, no material damage to liquid FLiBe blanket
• Difference of +/- ~10% for outboard/inboard
Material Layer Alphas (appm) 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
ZrH2 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
Average Outlet Temperature of 886 K
Inlet Temperature of 800 K
Inlet Velocity of 0.1 m/s
Surface Heat Flux
from Plasma
K
R
FLiBe Properties • Low electrical conductivity • Low toxicity • Twice density of water • Similar Cp to water • Similar viscosity to water • Melting point : 732 K
Peak Temperature of Blanket
COMSOL Turbulent Flow Model Determined VV Temperature Limits
Hea
t flu
x of
0.1
7 M
W/m
2 fr
om p
lasm
a
Fix
ed b
y bl
anke
t a
t 10
00 K
Channel length is the poloidal circumference of 10m
Beryllium Multiplier
FLiBe Channel
Inner VV
Outer VV
1.5 cm 2 cm 2 cm 1 cm
Inlet Temperature of 800 K Inlet Velocity of 4 m/s
Volumetric Neutron Heating (MW/m3)
10.8 5.4 22.5 9.8
Tungsten 1 mm
COMSOL Turbulent Flow Model Determined VV Temperature Limits
Outlet
Inlet
Unirradiated Inconel 718[1]
[1] Special Metals Corp., Pub. No. SMC-045, Sept. 2007
Outer VV
Beryllium Multiplier
FLiBe Channel
Inner VV
Peak temperature of VV
Unirradiated VV Survives Worst Case, Unmitigated Disruption
Von Mises Stresses (MPa)
• Yield Strength of unirradiated Inconel 718 at 1000K ≈ 1000 MPa • Von Mises stresses scale linearly with plasma current • Damage is localized to replaceable components
cmt
II
II
Bjf
Bjf
VV
p
phalo
VV
halohalo
5.2
6.0
4.0
cos1
MPa
Peak stress in VV
Integrated Design Indicates Tradeoffs in VV
1
1.02
1.04
1.06
1.08
1.1
1.12
1.14
1.16
1.18
1.2
0.9
1
1.1
1.2
1.3
1.4
1.5
1.6
1.7
1.8
0.5 1 1.5 2
Triti
um B
reed
ing
Ratio
Mec
hani
cal F
acto
r of S
afet
y*
Inner VV Thickness (cm)
* 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/cm2 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
• 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
ARC is a Dual-Purpose Fusion Facility: Both FNSF and Fusion Reactor
Questions?
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
0.00
5.00
10.00
15.00
20.00
25.00
30.00TF
Life
time
(yea
rs)
15 cm First Shield Material Effect on TF Lifetime
Tungsten
Tungsten Carbide
Beryllium
Graphite
ZrH2
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 recent1 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 databases1
• 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
fhalo = jhalo B =Ihalo 1+ cos ( )
2 R( ) tVV
jp
B0R0
R
e
fV = j BV = j BVeR
Ihalo = 0.4I pI = 0.6Ip
B0 = 9.2TBV = 0.8T tVV = 2.5cm
• Assumes a full, worst case, unmitigated disruption:
j BV
BV
B jhalo
j BV j
jhalo B
jhalo B
j
Turbulent Model
V = 4ms
T =1000KDH = 2 tChannel = 0.04m
= 5.9410 5 kgm s
e
4605 K[ ]T = 0.006 kg
m s
= 2413- 0.488 K 1 ( )*T( ) kgm3
=1925 kgm3
Re = DH
V =12833V = 51333
Turbulent
V = 0.1ms
T = 900KDH = 0.3m
= 0.01 kgm s
=1974 kgm3
Re = 59214V = 5921 Turbulent
VV Channel Flow Blanket Flow
Vacuum Vessel Temperature
kINC718 = 21 Wm K
xVV =1.5cm
qVV =0.3PheatingAVV
= 0.17MWm2
qVV = kINC718TVV = kINC718 TVV xVV
TVV =121KTflibe + TVV 1121K < Tmelt =1533K
