Prospects for Attractive Fusion Power Plants

Farrokh NajmabadiUniversity of California San Diego

18th KAIF/KNS WorkshopSeoul, Korea April 21, 2006

Electronic copy: http://aries.ucsd.edu/najmabadi/

ARIES Web Site: http:/aries.ucsd.edu/ARIES

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

ARIES-AT is an attractive vision for fusion with a reasonable extrapolation in physics & technology

Competitive cost of electricity (5c/kWh);

Steady-state operation; Low level waste; Public & worker safety; High availability.

A high-performance plasma should have a high power density & a low recirculating power fraction

Requirement: Establish and maintain the magnetic bottle

External magnets:Superconducting: size and costNormal conducting (e.g., copper): power consumption

Maintenance of plasma profiles (mainly plasma current)Inductive current-drive (transformer action): non-stationaryNon-inductive current drive by neutral beams, microwave ...: Inefficient

Requirement: Establish and maintain the magnetic bottle

External magnets:Superconducting: size and costNormal conducting (e.g., copper): power consumption

Maintenance of plasma profiles (mainly plasma current)Inductive current-drive (transformer action): non-stationaryNon-inductive current drive by neutral beams, microwave ...: Inefficient

Key parameters: Plasma (ratio of plasma pressure to magnetic pressure)

Non-dimensional parameter N is a measure of plasma performance

Current-drive power PCD

Key parameters: Plasma (ratio of plasma pressure to magnetic pressure)

Non-dimensional parameter N is a measure of plasma performance

Current-drive power PCD

Our vision of a fusion system in 1980s was a large pulsed device. Non-inductive current drive is inefficient.

Some important achievements in ~1990: 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 ~1990: 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

Reverse Shear Regime Excellent match between bootstrap & equilibrium current profile at high ARIES-RS: N= 4.8, =5%, Pcd=81 MW (achieves ~5 MW/m2 peak wall

loading.) ARIES-AT (advanced physics & technology): N= 5.4, =9%, Pcd=36 MW

(high is used to reduce peak field at magnet)

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

Requirement: Plasma should be surrounded by a blanket containing Li

Plasma: D + T He + n

Blanket: n + 6Li T + He

DT fusion turns its waste (neutrons) into fuel!

Through careful design, only a small fraction of neutrons are absorbed in structure and induce radioactivityRad-waste depends on the choice of material: Low-activation materialRad-waste generated in DT fusion is similar to advanced fuels (D-3He)For liquid coolant/breeders (e.g., Li, LiPb), most of fusion energy (carried

by neutrons) is directly deposited in the coolant simplifying energy recovery

Issue: Large flux of neutrons through the first wall and blanket:Need to develop radiation-resistant, low-activation material:

Ferritic steels, Vanadium alloys, SiC composites

Requirement: Plasma should be surrounded by a blanket containing Li

Plasma: D + T He + n

Blanket: n + 6Li T + He

DT fusion turns its waste (neutrons) into fuel!

Through careful design, only a small fraction of neutrons are absorbed in structure and induce radioactivityRad-waste depends on the choice of material: Low-activation materialRad-waste generated in DT fusion is similar to advanced fuels (D-3He)For liquid coolant/breeders (e.g., Li, LiPb), most of fusion energy (carried

by neutrons) is directly deposited in the coolant simplifying energy recovery

Issue: Large flux of neutrons through the first wall and blanket:Need to develop radiation-resistant, low-activation material:

Ferritic steels, Vanadium alloys, SiC composites

DT Fusion requires a T breeding blanket

D + 6Li He + He

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.