Transcript Addressing Fusion Power Issues Presented by R.D. Stambaugh Harnessing Fusion Power

Addressing Fusion Power Issues
Presented by
R.D. Stambaugh
PERSISTENT SURVEILLANCE FOR
PIPELINE PROTECTION AND THREAT INTERDICTION
Harnessing Fusion Power
ReNeW Workshop
March 3, 2009
UCLA
35906/RDS/rs
13806/TST/rs
The Fusion Development Facility Mission:
Develop Fusion’s Energy Applications
• FDF will:
– Close the fusion fuel cycle
– Show electricity production from fusion.
– Develop high temperature blankets for high efficiency electricity
production.
– Show hydrogen production from fusion.
– Provide a materials irradiation facility to develop fusion materials.
• By using conservative Advanced Tokamak physics to run steady-state
and produce 100-250 MW fusion power
– Modest energy gain (Q<5)
– Continuous operation for 30% of a year in 2 weeks periods
– Test materials with high neutron fluence (3-8 MW-yr/m2)
– Further develop all elements of Advanced Tokamak physics,
qualifying them for an advanced performance DEMO
• With ITER and IFMIF, provide the basis for a fusion DEMO Power Plant
35906/RDS/rs
FDF is Viewed as a Direct Follow-on of DIII-D (50% larger)
and Alcator Cmod, Using Their Construction Features
DIII-D
FDF
• Plate constructed
copper TF Coil
which enables..
• TF Coil joint for complete
dissasembly and
maintenance
• OH Coil wound on the TF
Coil to maximize Voltseconds
• High elongation, high
triangularity double null
plasma shape for high
gain, steady-state
• Red blanket produces
net Tritium
35906/RDS/rs
FDF Dimensions for Reference
PF1 1.2 X 0.55
TF WEDGE
TF CENTER POST
PF2
0.65 X 0.65
BLANKET 0.5 THK
PF3
0.65 X 0.65
OHMIC
HEATING COIL
7.15
R
1.42
R
2.49
R
1.78
TF OUTER
VERTICALS
R 1.2
R 4.35
35906/RDS/rs
R 5.72
The U.S. Fusion Nuclear Science Community
Suggested an Aggressive Phased Research Plan
35906/RDS/rs
To Show Fusion Can Close its Fuel Cycle,
FDF Will Demonstrate Efficient Net Tritium Production
• FDF will produce 0.4–1.3 kg of Tritium per year at its nominal
duty factor of 0.3
• This amount should be sufficient for FDF and can build the
T supply needed for DEMO
For TBR 1.2
Tritium Produced Annually
5.00
4.50
4.00
3.50
kg
3.00
125 MW
2.50
250 MW
400 MW
2.00
1.50
1.00
0.50
0.00
0
0.2
0.4
0.6
Duty Factor
35906/RDS/rs
0.8
1
1.2
Port Sites Enable Nuclear and Materials Science.
DIII-D size neutral beams
- 3 Co 120 keV, rotation
- 1 Counter, 80 keV
for QH mode edge
Off-axis current profile control
- ECCD (170 GHz)
- Lower Hybrid
- NBCD
Port blanket sites for fusion
nuclear technology
development
Port blanket sites for fusion
materials development
35906/RDS/rs
FDF will Motivate the Needed, Large, Supporting
Fusion Nuclear Science Program
On Test Specimens and Components,
• Materials compositions
• Activation and transmutation
• Materials properties (irradiated)
• Thermo-hydraulics
• Thermal expansion and stress
• Mechanical and EM stresses
• Tritium breeding and retention
• Solubility, diffusivity, permeation
• Liquid metals crossing magnetic fields
• Coolant properties
• Chemistry
• and more…..
35906/RDS/rs
8
FDF is a Materials Irradiation Research Facility
• Provides up to 80 dpa of DT fusion neutron
irradiation in controlled environment
materials test ports for:
– First wall chamber materials
– Structural materials
– Breeders
– Neutron multipliers
– Tritium permeation barriers
– Composites
– Electrical and thermal insulators
• Materials compatibility tests in an integrated
tokamak environment
– Flow channel inserts for DCLL blanket
option
– Chamber components and diagnostics
development
35906/RDS/rs
Materials Which Could Be Irradiation Tested in FDF
to 60 dpa
PFC surface materials:
W, Mo…etc.
Joining materials: W to RAFM, W to ODFS,
etc.
Engineered materials: BW, Low-Z+UFG
refractory…etc.
Flow-through materials: e.g. C, B
Structural materials:
Different RAFM alloys: e.g. F82H,
EUROFER…etc.
Flow channel inserts for DCLL:
SiC/SiC composite, SiC-foam
MHD insulation for self-cooled options:
Y2O3, Er2O3, AlN, sandwich layers…etc.
Advanced structural materials:
Different ODFS alloys, W-Re alloys, W-TiC
alloys, Mo-alloys, SiC/SiC…etc
Structural joint materials:
35906/RDS/rs
Solid breeder materials:
Li2O, Li2TiO3, Li4SiO4
Liquid breeder materials:
Pb17Li, FLiBe…others.
Neutron multipliers:
Be, BeTi…etc.
Tritium barriers:
Al2O3…etc.
Low flux tests:
Superconducting materials for low
and high temp. superconductors
Electrical and thermal
insulator…etc.
Shielding materials:
SS, FS, W, B…etc.
FDF Will Develop Blankets for Fusion Electric Power
• Fusion electric blankets require
– High temperature (500-700 0C) heat extraction
– Complex neutronics issues
– Tritium breeding ratio > 1.0
– Chemistry effects (hot, corrosive, neutrons)
– Environmentally attractive materials
– High reliability, (disruptions, off-normal events)
• Fusion blanket development requires testing
– Solid breeders (3), Liquid breeders (2)
– Various Coolants (2)
– Advanced, Low Activation, Structural materials (2)
• Desirable capabilities of a development facility
– 1–2 MW/m2 14 MeV neutron flux
– 10 m2 test area, relevant gradients(heat, neutrons)
– Continuous on time of 1-2 weeks
– Integrated testing with fluence 6 MW-yr/m2
• FDF can deliver all the above testing requirements
– Test two blankets every two years
– In ten years, test 10 blanket approaches
Produce 300 kW electricity from one port blanket
35906/RDS/rs
Teams of Universities, Labs, and Industry Will Work
with the Site to Field Test Blanket Modules on FDF
• Helium-cooled Lithium-Lead blanket (1 type)
• Dual-Coolant (He and LiPb) type LithiumLead (DFLL and DCLL) blankets (2 types)
• Dual-Coolant (He and LiPb) Lithium-Lead
Ceramic Breeder (LLCB) blanket (1 type)
• Helium-cooled Ceramic/Beryllium blanket
(3 designs)
• Water-cooled Ceramic/Beryllium blanket
(1 type)
These are 7 test blanket options being evaluated by the ITER
TBM program, and with detailed variations and gradual
improvements of preferred concepts, judicious selection will be
needed to focus on the selection of ~ 10 blanket concept
designs to be tested in FDF
35906/RDS/rs
FDF Will Develop Hydrogen Production from Fusion
7.1 MPa
450ЎC
222 kg/s
Sulfur-Iodine Cycle
CIRC
H2
LOW
TEMP
HX
MHR 600 MW(t)
CIRC
HI
DECOMPOSER
I2
HI
636ЎC
CHEMICAL
REACTOR
IHX
H2O
O2
H2SO4
925ЎC
HIGH
TEMP
HX
H2SO4
DECOMPOSER
SO2 + H2O
+ O2
7.0 MPa
950ЎC
320 kg/s
Requires 900-1000˚C Blankets
Perhaps one metric ton per
week from one port plug
35906/RDS/rs
ITER, FDF, and IFMIF Solve the Gap Issues for DEMO
35906/RDS/rs
FDF Makes Major Contributions to Almost All Gaps
Identified by the FESAC Planning Panel
35906/RDS/rs
FDF Supports a Variety of Operating Modes to Support
Nuclear Science and Advanced Tokamak toward DEMO
Wall Load
2 MW/m2
A
a
Ro
Elongation
Fusion Power
Plant Power
Pn/Awall
Qplasma
BetaT
BetaN
fbs
Pcd
Paux
Ip
Bo
q
Ti(0)
n(0)
nbar/nGR
Zeff
W
TauE
HITER98Y2
PTotal/R
Peak Heat Flux
m
m
MW
MW
MW/m2
mT/MA
MW
MW
MA
T
keV
E20/m3
MJ
sec
MW/m
MW/m2
35906/RDS/rs
3.5
0.71
2.49
2.31
246
507
2.0
4.2
5.8%
3.7
60%
59
59
6.7
6.0
5.0
19
3.0
0.57
2.1
70
0.6
1.60
43
8.2
1.0 MW/m2, High Gain
Very
Very
ITER-SS
ARIES-AT
Lower B, fbs Inductive
Advanced
Advanced
3.5
3.5
3.5
3.5
3.4
4
0.71
0.71
0.71
0.71
1.85
1.30
2.49
2.49
2.49
2.49
6.35
5.20
2.31
2.31
2.31
2.31
1.85
2.20
123
231
301
401
356
1755
362
395
482
536
1.0
1.9
2.5
3.3
0.5
4.8
2.5
11.5
4.5
6.1
6.0
45.0
7.6%
9.2%
7.9%
7.4%
2.8%
9.2%
3.7
3.3
4.5
4.5
3.0
5.4
46%
30%
65%
70%
48%
91%
50
20
65
66
35
50
20
67
66
59
36
6.5
9.3
6.8
7.0
9.0
12.8
4.4
4.7
5.4
6.0
5.2
5.8
3.8
2.8
4.4
4.8
5.3
3.7
20
16
18
18
19
31
2.0
3.5
3.5
4.1
0.7
2.9
0.40
0.47
0.66
0.74
0.82
0.96
2.1
2.1
2.1
2.1
2.1
1.7
50
67
77
89
287
640
0.7
1.0
0.6
0.6
3.1
2.0
1.60
1.36
1.59
1.60
1.57
1.40
30
27
51
59
21
74
6.1
3.8
9.3
10.2
10.0
9.3
FDF Will Be a Necessary Progress Step to DEMO in Dealing
With Off-Normal Events
• Goals
– Operate for 107 seconds per year, duty factor 0.3.
– Run two weeks straight without disruption
– Only one unmitigated disruption per year
• Disruption Strategy
– Real-time stability calculations in the control loop.
– Active instability avoidance and suppression RWMs, NTMs
– Control system good enough to initiate soft
shutdowns and limit firing the disruption mitigation
system more than 20 times per year.
– Disruption mitigation system 99% reliable.
• ELM Suppression
– Resonant magnetic perturbation coils
– QH-mode
35906/RDS/rs
Reliability, Availability, Maintainability, and Inspectability
Have Dominated the FDF Design Concept
• It is the main reason the machine is based on a jointed copper coil
– To enable planned changes in the entire blanket structure
– So the entire machine can be taken apart readily and any component
remotely serviced.
– So the nuclear components can be accessible by simple crane lift.
• The machine must be reliable to achieve continuous two week operation and
the availability to achieve a duty factor 0.3 on a year.
• It must be maintainable because it is a research environment and problems
must be repairable and the blankets changeable.
• Inspection of the components is an integral part of the research; we need to
find out what is happening to all these components.
• The FDF Program will be a test bed for learning how to engineer reliable first
wall/blanket structures and to gain first information on reliability growth.
35906/RDS/rs
The Baseline Maintenance Scheme is
Toroidally Continuous Blanket Structures
Remove
• Upper sections of TF
• Divertor coil
• Top of vacuum vessel
Access to blanket structure
obtained
• Blanket segments removed as
toroidally continuous rings
Benefits
• Blankets strong for EM loads
• Toroidal alignment assured
Difficulties
• Provision of services (coolants)
to blanket rings near the
midplane through blankets
above
35906/RDS/rs
Second option: Vertical Removal of Poloidal
Blanket Wedge Sectors
Features:
• Divertor coil located outside TF
Process:
• Lift off Divertor coil
• TF upper section(s) removed
• Remove top vessel section
• Blanket sector removed vertically
Benefits
• Access for localized repair
• Blankets of different types could be
installed
• Coolant services from top and bottom
localized to each sector
Difficulty
• Alignment of modules critical
35906/RDS/rs
A New DT Burning Plasma Facility Should Be Built in the
US to provide a Fusion Nuclear Science “Laboratory.”
•
•
•
•
•
•
Develop fusion’s energy applications.
Close the fusion fuel cycle.
Develop blankets for fusion electric power.
Develop hydrogen production from fusion.
Address nearly all gaps Identified by FESAC.
Motivate the needed, large, supporting fusion nuclear science
program
• Provide a materials irradiation and research facility
• FDF should be the next major U.S. facility running in parallel
with ITER
35906/RDS/rs