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Towards Attractive Fusion Power Plants
Farrokh NajmabadiUniversity of California San Diego
Presented atKorean National Fusion Research CenterDaejon, Korea April 20, 2006
Electronic copy: http://aries.ucsd.edu/najmabadi/
ARIES Web Site: http:/aries.ucsd.edu/ARIES
The ARIES Team Has Examined Several Magnetic Fusion Concept as Power Plants in the Past 15 Years
• TITAN reversed-field pinch (1988)
• ARIES-I first-stability tokamak (1990)
• ARIES-III D-3He-fueled tokamak (1991)
• ARIES-II and -IV second-stability tokamaks (1992)
• Pulsar pulsed-plasma tokamak (1993)
• SPPS stellarator (1994)
• Starlite study (1995) (goals & technical requirements for power plants & Demo)
• ARIES-RS reversed-shear tokamak (1996)
• ARIES-ST spherical torus (1999)
• Fusion neutron source study (2000)
• ARIES-AT advanced technology and advanced tokamak (2000)
• ARIES-IFE assessment of Inertial Fusion chambers, targets, and drivers (2003)
• ARIES-CS compact stellarator (2006)
Framework:Assessment Based on Attractiveness & Feasibility
Periodic Input fromEnergy Industry
Goals and Requirements
Scientific & TechnicalAchievements
Evaluation Based on Customer Attributes
Attractiveness
Characterizationof Critical Issues
Feasibility
Projections andDesign Options
Balanced Assessment ofAttractiveness & Feasibility
No: RedesignR&D Needs and
Development Plan
Yes
Elements of the Case for Fusion Power Were Developed through Interaction with Representatives of U.S. Electric Utilities and Energy Industry
Have an economically competitive life-cycle cost of electricity
Gain Public acceptance by having excellent safety and environmental characteristics No disturbance of public’s day-to-day activities No local or global atmospheric impact No need for evacuation plan No high-level waste Ease of licensing
Reliable, available, and stable as an electrical power source Have operational reliability and high availability Closed, on-site fuel cycle High fuel availability Capable of partial load operation Available in a range of unit sizes
Fu
sion
ph
ysic
s &
te
chn
olog
y
Low-activation material
Framework:Assessment Based on Attractiveness & Feasibility
Periodic Input fromEnergy Industry
Goals and Requirements
Scientific & TechnicalAchievements
Evaluation Based on Customer Attributes
Attractiveness
Characterizationof Critical Issues
Feasibility
Projections andDesign Options
Balanced Assessment ofAttractiveness & Feasibility
No: RedesignR&D Needs and
Development Plan
Yes
Our vision of a fusion system in 1980s was a large pulsed device. Non-inductive current drive is inefficient.
Some important achievements in 1980s: Experimental demonstration of bootstrap current; Development of ideal MHD codes that agreed with experimental results.
Development of steady-state power plant concepts (ARIES-I and SSTR) based on the trade-off of bootstrap current fraction and plasma ARIES-I: N= 2.9, =2%, Pcd=230 MW
Our vision of a fusion system in 1980s was a large pulsed device. Non-inductive current drive is inefficient.
Some important achievements in 1980s: Experimental demonstration of bootstrap current; Development of ideal MHD codes that agreed with experimental results.
Development of steady-state power plant concepts (ARIES-I and SSTR) based on the trade-off of bootstrap current fraction and plasma ARIES-I: N= 2.9, =2%, Pcd=230 MW
A dramatic change occurred in 1990: Introduction of Advanced Tokamak
Last decade: Reverse Shear Regime Excellent match between bootstrap & equilibrium current profile at high Requires wall stabilization (Resistive-wall modes). Internal transport barrier.
Framework:Assessment Based on Attractiveness & Feasibility
Periodic Input fromEnergy Industry
Goals and Requirements
Scientific & TechnicalAchievements
Evaluation Based on Customer Attributes
Attractiveness
Characterizationof Critical Issues
Feasibility
Projections andDesign Options
Balanced Assessment ofAttractiveness & Feasibility
No: RedesignR&D Needs and
Development Plan
Yes
Increase Power Density
Directions for Improvement
What we pay for,VFPC
r
Power density, 1/Vp
r > r ~ r <
Improvement “saturates” at ~5 MW/m2 peak wall loading (for a 1GWe plant).
A steady-state, first stability device with Nb3Sn technology has a power density about 1/3 of this goal.
Big Win Little
Gain
Decrease Recirculating Power Fraction Improvement “saturates” about Q ~ 40. A steady-state, first stability device with Nb3Sn Tech.
has a recirculating fraction about 1/2 of this goal.
High-Field Magnets
ARIES-I with 19 T at the coil (cryogenic).
Advanced SSTR-2 with 21 T at the coil (HTS).
High bootstrap, High 2nd Stability: ARIES-II/IV Reverse-shear: ARIES-RS,
ARIES-AT, A-SSRT2
Reduced COE mainly due to advanced technology
Approaching COE insensitive of power density
Evolution of ARIES Designs
1st Stability,
Nb3Sn Tech.
ARIES-I’
Major radius (m) 8.0
) 2% (2.9)
Peak field (T) 16
Avg. Wall Load (MW/m2) 1.5
Current-driver power (MW) 237
Recirculating Power Fraction 0.29
Thermal efficiency 0.46
Cost of Electricity (c/kWh) 10
Reverse Shear
Option
High-Field
Option
ARIES-I
6.75
2% (3.0)
19
2.5
202
0.28
0.49
8.2
ARIES-RS
5.5
5% (4.8)
16
4
81
0.17
0.46
7.5
ARIES-AT
5.2
9.2% (5.4)
11.5
3.3
36
0.14
0.59
5
ARIES-I Introduced SiC Composites as A High-Performance Structural Material for Fusion
Excellent safety & environmental characteristics (very low activation and very low afterheat).
High performance due to high strength at high temperatures (>1000
oC).
Large world-wide program in SiC: New SiC composite fibers with proper
stoichiometry and small O content. New manufacturing techniques based
on polymer infiltration or CVI result in much improved performance and cheaper components.
Recent results show composite thermal conductivity (under irradiation) close to 15 W/mK which was used for ARIES-I.
Continuity of ARIES research has led to the progressive refinement of research
ARIES-I:
• SiC composite with solid breeders
• Advanced Rankine cycle
ARIES-I:
• SiC composite with solid breeders
• Advanced Rankine cycle
Starlite & ARIES-RS:
• Li-cooled vanadium
• Insulating coating
Starlite & ARIES-RS:
• Li-cooled vanadium
• Insulating coating
ARIES-ST:
• Dual-cooled ferritic steel with SiC inserts
• Advanced Brayton Cycle at 650 oC
ARIES-ST:
• Dual-cooled ferritic steel with SiC inserts
• Advanced Brayton Cycle at 650 oC
ARIES-AT:
• LiPb-cooled SiC composite
• Advanced Brayton cycle with = 59%
ARIES-AT:
• LiPb-cooled SiC composite
• Advanced Brayton cycle with = 59%
Many issues with solid breeders; Rankine cycle efficiency saturated at high temperature
Max. coolant temperature limited by maximum structure temperature
High efficiency with Brayton cycle at high temperature
Advanced Brayton Cycle Parameters Based on Present or Near Term Technology Evolved with Expert Input from General Atomics*
Key improvement is the development of cheap, high-efficiency recuperators.
RecuperatorIntercooler 1Intercooler 2
Compressor 1
Compressor 2Compressor 3
HeatRejection
HX
Wnet
Turbine
Blanket
IntermediateHX
5'
1
22'
38
9
4
7'9'
10
6
T
S
1
2
3
4
5 6 7 8
9 10
Divertor
LiPbBlanketCoolant
He DivertorCoolant
11
11
Originally developed for ARIES-ST, further developed by EU (FZK). Typically, the coolant outlet temperature is limited to the max. operating
temperature of structural material (550oC for ferritic steels).
A coolant outlet temperature of 700oC is achieved by using a coolant/breeder
(LiPb), cooling the structure by He gas, and SiC insulator lining PbLi channel for thermal and electrical insulation.
Originally developed for ARIES-ST, further developed by EU (FZK). Typically, the coolant outlet temperature is limited to the max. operating
temperature of structural material (550oC for ferritic steels).
A coolant outlet temperature of 700oC is achieved by using a coolant/breeder
(LiPb), cooling the structure by He gas, and SiC insulator lining PbLi channel for thermal and electrical insulation.
ARIES-ST Featured a High-Performance Ferritic Steel Blanket
Simple, low pressure design with SiC structure and LiPb coolant and breeder.
Outboard blanket & first wall
ARIES-AT2: SiC Composite Blankets
Simple manufacturing technique.
Very low afterheat.
Class C waste by a wide margin.
LiPb-cooled SiC composite divertor is capable of 5 MW/m2 of heat load.
Innovative design leads to high LiPb outlet temperature (~1,100oC) while keeping SiC structure temperature below 1,000oC leading to a high thermal efficiency of ~ 60%.
Innovative Design Results in a LiPb Outlet Temperature of 1,100oC While Keeping SiC Temperature Below 1,000oC
• Two-pass PbLi flow, first pass to cool SiCf/SiC box second pass to superheat PbLi
q''plasma
Pb-17Li
q'''LiPb
Out
q''back
vback
vFW
Poloidal
Radial
Inner Channel
First Wall Channel
SiC/SiCFirst Wall SiC/SiC Inner Wall
700
800
900
1000
1100
1200800
900
1000
1100
1200
1
2
3
4
5
6
00.020.040.060.080.1
00.020.040.060.080.1
Radial distance (m)
Poloidaldistance(m)
SiC/SiC
Pb-17Li
Bottom
Top
PbLi Outlet Temp. = 1100 °C
Max. SiC/PbLi Interf. Temp. = 994 °C
Max. SiC/SiC Temp. = 996°C
PbLi Inlet Temp. = 764 °C
The ARIES-AT Utilizes An Efficient Superconducting Magnet Design
On-axis toroidal field: 6 T
Peak field at TF coil: 11.4 T
TF Structure: Caps and straps support loads without inter-coil structure;
On-axis toroidal field: 6 T
Peak field at TF coil: 11.4 T
TF Structure: Caps and straps support loads without inter-coil structure;
Superconducting Material Either LTC superconductor (Nb3Sn
and NbTi) or HTC Structural Plates with grooves for
winding only the conductor.
Superconducting Material Either LTC superconductor (Nb3Sn
and NbTi) or HTC Structural Plates with grooves for
winding only the conductor.
Use of High-Temperature Superconductors Simplifies the Magnet Systems
HTS does offer operational advantages: Higher temperature operation
(even 77K), or dry magnets Wide tapes deposited directly
on the structure (less chance of energy dissipating events)
Reduced magnet protection concerns
HTS does offer operational advantages: Higher temperature operation
(even 77K), or dry magnets Wide tapes deposited directly
on the structure (less chance of energy dissipating events)
Reduced magnet protection concerns
Inconel strip
YBCO Superconductor Strip Packs (20 layers each)
8.5 430 mm
CeO2 + YSZ insulating coating(on slot & between YBCO layers)
Epitaxial YBCOEpitaxial YBCO Inexpensive manufacturing
would consist of layering HTS on structural shells with minimal winding!
Epitaxial YBCOEpitaxial YBCO Inexpensive manufacturing
would consist of layering HTS on structural shells with minimal winding!
Modular sector maintenance enables high availability
Full sectors removed horizontally on rails Transport through maintenance corridors
to hot cells Estimated maintenance time < 4 weeks
Full sectors removed horizontally on rails Transport through maintenance corridors
to hot cells Estimated maintenance time < 4 weeks
ARIES-AT elevation view
Framework:Assessment Based on Attractiveness & Feasibility
Periodic Input fromEnergy Industry
Goals and Requirements
Scientific & TechnicalAchievements
Evaluation Based on Customer Attributes
Attractiveness
Characterizationof Critical Issues
Feasibility
Projections andDesign Options
Balanced Assessment ofAttractiveness & Feasibility
No: RedesignR&D Needs and
Development Plan
Yes
Our Vision of Magnetic Fusion Power Systems Has Improved Dramatically in the Last Decade, and Is Directly Tied to Advances in Fusion Science & Technology
Estimated Cost of Electricity (c/kWh)
0
2
4
6
8
10
12
14
Mid 80'sPhysics
Early 90'sPhysics
Late 90's Physics
AdvancedTechnology
Major radius (m)
0
1
2
3
4
5
6
7
8
9
10
Mid 80's Pulsar
Early 90'sARIES-I
Late 90'sARIES-RS
2000 ARIES-AT
Approaching COE insensitive of power density High Thermal Efficiency
High is used to lower magnetic field
10-7
10-6
10-5
10-4
10-3
10-2
10-1
100
101
104 105 106 107 108 109 1010 1011
ARIES-STARIES-RS
Act
ivit
y (
Ci/
W th)
Time Following Shutdown (s)
1 mo 1 y 100 y1 d
Radioactivity Levels in Fusion Power PlantsAre Very Low and Decay Rapidly after Shutdown
After 100 years, only 10,000 Curies
of radioactivity remain in the
585 tonne ARIES-RS fusion core.
After 100 years, only 10,000 Curies
of radioactivity remain in the
585 tonne ARIES-RS fusion core.
SiC composites lead to a very low activation and afterheat.
All components of ARIES-AT qualify for Class-C disposal under NRC and Fetter Limits. 90% of components qualify for Class-A waste.
SiC composites lead to a very low activation and afterheat.
All components of ARIES-AT qualify for Class-C disposal under NRC and Fetter Limits. 90% of components qualify for Class-A waste.
Ferritic SteelVanadium
Fusion Core Is Segmented to Minimize the Rad-Waste
Only “blanket-1” and divertors are replaced every 5 years
Only “blanket-1” and divertors are replaced every 5 years
Blanket 1 (replaceable)
Blanket 2 (lifetime)
Shield (lifetime)
Generated radioactivity waste is reasonable
0
50
100
150
200
250
300
350
400
Blanket Shield VacuumVessel
Magnets Structure Cryostat
Cu
mu
lati
ve
Co
mp
ac
ted
Wa
ste
Vo
lum
e (
m3
)
1270 m3 of Waste is generated after 40 full-power year (FPY) of operation (~50 years) Coolant is reused in other power plants 29 m3 every 4 years (component replacement) 993 m3 at end of service
Equivalent to ~ 30 m3 of waste per FPY Effective annual waste can be reduced by increasing plant service life.
1270 m3 of Waste is generated after 40 full-power year (FPY) of operation (~50 years) Coolant is reused in other power plants 29 m3 every 4 years (component replacement) 993 m3 at end of service
Equivalent to ~ 30 m3 of waste per FPY Effective annual waste can be reduced by increasing plant service life.
90% of waste qualifies for Class A disposal
90% of waste qualifies for Class A disposal
Framework:Assessment Based on Attractiveness & Feasibility
Periodic Input fromEnergy Industry
Goals and Requirements
Scientific & TechnicalAchievements
Evaluation Based on Customer Attributes
Attractiveness
Characterizationof Critical Issues
Feasibility
Projections andDesign Options
Balanced Assessment ofAttractiveness & Feasibility
No: RedesignR&D Needs and
Development Plan
Yes
Advances in plasma physics has led to a dramatic improvement in our vision of fusion systems
Attractive visions for tokamak exist. The main question is to what extent the advanced tokamak
modes can be achieved in a burning plasma (e.g., ITER): What is the achievable N (macroscopic stability)
Can the necessary pressure profiles realized in the presence of strong heating (microturbulence & transport)
Attractive visions for tokamak exist. The main question is to what extent the advanced tokamak
modes can be achieved in a burning plasma (e.g., ITER): What is the achievable N (macroscopic stability)
Can the necessary pressure profiles realized in the presence of strong heating (microturbulence & transport)
Pace of “Technology” research has been considerably slower than progress in plasma physics.
Advanced technologies have a dramatic impact on attractiveness of fusion. Considerable more technology R&D is needed.
Pace of “Technology” research has been considerably slower than progress in plasma physics.
Advanced technologies have a dramatic impact on attractiveness of fusion. Considerable more technology R&D is needed.