Fusion Nuclear Technology:
Principles and Challenges
Mohamed Abdou (web: http://www.fusion.ucla.edu/abdou/)
Distinguished Professor of Engineering and Applied Science
Director, Center for Energy Science and Technology (CESTAR)
(http://www.cestar.seas.ucla.edu/)
Director, Fusion Science and Technology Center (http://www.fusion.ucla.edu/)
University of California, Los Angeles (UCLA)
Seminar at Grenoble, France
March 23, 2007
1
Incentives for Developing Fusion
Fusion powers the Sun and the stars
 It is now within reach for use on Earth
In the fusion process lighter elements are “fused” together, making
heavier elements and producing prodigious amounts of energy
Fusion offers very attractive features:
 Sustainable energy source
(for DT cycle; provided that Breeding Blankets are successfully
developed)
 No emission of Greenhouse or other polluting gases
 No risk of a severe accident
 No long-lived radioactive waste
Fusion energy can be used to produce electricity and hydrogen, and
for desalination
2
ITER
• The World is about to construct the next
step in fusion development, a device
called ITER
• ITER will demonstrate the scientific and
technological feasibility of fusion energy
for peaceful purposes
• ITER will produce 500 MW of fusion power
• Cost, including R&D, is 15 billion dollars
3
ITER Design - Main Features
Central
Solenoid
Outer Intercoil
Structure
Blanket
Module
Vacuum Vessel
Cryostat
Toroidal Field Coil
Port Plug (IC Heating
Poloidal Field Coil
Divertor
Machine Gravity Supports
Torus Cryopump
4
ITER is a collaborative effort among Europe,
Japan, US, Russia, China and South Korea
5
Fusion Nuclear Technology (FNT)
Fusion Power & Fuel Cycle Technology
FNT Components from the edge of the
Plasma to TF Coils (Reactor “Core”)
1. Blanket Components
2. Plasma Interactive and High Heat Flux
Components
a. divertor, limiter
b. rf antennas, launchers, wave guides, etc.
3. Vacuum Vessel & Shield Components
Other Components affected by the
Nuclear Environment
4. Tritium Processing Systems
5. Instrumentation and Control Systems
6. Remote Maintenance Components
7. Heat Transport and Power Conversion
Systems
6
The Deuterium-Tritium (D-T) Cycle
• World Program is focused on the D-T cycle (easiest to
ignite):
D + T → n + α + 17.58 MeV
• The fusion energy (17.58 MeV per reaction) appears as
Kinetic Energy of neutrons (14.06 MeV) and alphas (3.52
MeV)
• Tritium does not exist in nature! Decay half-life is 12.3 years
(Tritium must be generated inside the fusion system to
have a sustainable fuel cycle)
• The only possibility to adequately breed tritium is through
neutron interactions with lithium
– Lithium, in some form, must be used in the fusion system
7
Tritium Breeding
Li-6(n,alpha)t and Li-7(n,n,alpha)t Cross-Section
1000
Natural lithium contains
7.42% 6Li and 92.58% 7Li.
100
6Li
(n,a) t
Li-6(n,a) t
Li-7(n,na)t
10
6
Li  n  t  a  4.78MeV
7
Li  n  t  a  n  2.47 MeV
1
The 7Li(n;n’a)t reaction is a
threshold reaction and
requires an incident neutron
energy in excess of 2.8 MeV.
0.1
7Li
(n;n’a) t
0.01
1
10
100
1000
10
4
10
Neutron Energy (eV)
5
10
6
10
7
8
Blanket (including first wall)
Blanket Functions:
A. Power Extraction
–
Convert kinetic energy of neutrons and secondary gamma rays into heat
–
Absorb plasma radiation on the first wall
–
Extract the heat (at high temperature, for energy conversion)
B. Tritium Breeding
–
Tritium breeding, extraction, and control
–
Must have lithium in some form for tritium breeding
C. Physical Boundary for the Plasma
–
Physical boundary surrounding the plasma, inside the vacuum vessel
–
Provide access for plasma heating, fueling
–
Must be compatible with plasma operation
–
Innovative blanket concepts can improve plasma stability and confinement
D. Radiation Shielding of the Vacuum Vessel
9
Blanket Concepts
(many concepts proposed worldwide)
A.
B.
Solid Breeder Concepts
–
Always separately cooled
–
Solid Breeder: Lithium Ceramic (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3)
–
Coolant: Helium or Water
Liquid Breeder Concepts
Liquid breeder can be:
a) Liquid metal (high conductivity, low Pr): Li, or 83Pb 17Li
b) Molten salt (low conductivity, high Pr): Flibe (LiF)n · (BeF2),
Flinabe (LiF-BeF2-NaF)
B.1. Self-Cooled
–
Liquid breeder is circulated at high enough speed to also serve as coolant
B.2. Separately Cooled
–
A separate coolant is used (e.g., helium)
–
The breeder is circulated only at low speed for tritium extraction
B.3. Dual Coolant
–
FW and structure are cooled with separate coolant (He)
–
Breeding zone is self-cooled
10
A Helium-Cooled Li-Ceramic Breeder Concept: Example
Material Functions
• Beryllium (pebble bed) for
neutron multiplication
• Ceramic breeder (Li4SiO4,
Li2TiO3, Li2O, etc.) for tritium
breeding
• Helium purge (low pressure)
to remove tritium through
the “interconnected
porosity” in ceramic breeder
• High pressure Helium
cooling in structure (ferritic
steel)
Several configurations exist (e.g. wall parallel or “head on”
breeder/Be arrangements)
11
Examples of solid breeder blankets that utilize immobile solid lithium
ceramic breeder and Be multiplier for tritium self-sufficiency
EU HCPB DEMO Module
JA HCSB Module
Tritium Breeder
Li2TiO3, Li2O (<2mm)
Neutron Multiplier
Be, Be12Ti (<2mm)
First Wall
(RAFS, F82H)
Breeder Unit to be inserted into
the space between the grid plates
phase I ARIES-CS HCCB
ODS +FS
12
Solid Breeder Blanket Issues
 Tritium self-sufficiency
 Breeder/Multiplier/structure interactive effects under
nuclear heating and irradiation
 Tritium inventory, recovery and control; development of
tritium permeation barriers
 Effective thermal conductivity, interface thermal
conductance, thermal control
 Allowable operating temperature window for breeder
 Failure modes, effects, and rates
 Mass transfer
 Temperature limits for structural materials and coolants
 Mechanical loads caused by major plasma disruption
 Response to off-normal conditions
13
Liquid Breeders
•
Many liquid breeder concepts exist, all of which have
key feasibility issues. Selection can not prudently be
made before additional R&D results become available.
•
Type of Liquid Breeder: Two different classes of
materials with markedly different issues.
a)
Liquid Metal: Li, 83Pb 17Li
High conductivity, low Pr number
Dominant issues: MHD, chemical reactivity for Li, tritium
permeation for LiPb
b)
Molten Salt: Flibe (LiF)n · (BeF2), Flinabe (LiF-BeF2-NaF)
Low conductivity, high Pr number
Dominant Issues: Melting point, chemistry, tritium control
14
Liquid Breeder Blanket Concepts
1.
Self-Cooled
–
Liquid breeder circulated at high speed to serve as coolant
–
Concepts: Li/V, Flibe/advanced ferritic, flinabe/FS
2.
Separately Cooled
–
A separate coolant, typically helium, is used. The breeder is
circulated at low speed for tritium extraction.
–
Concepts: LiPb/He/FS, Li/He/FS
3.
Dual Coolant
–
First Wall (highest heat flux region) and structure are cooled
with a separate coolant (helium). The idea is to keep the
temperature of the structure (ferritic steel) below 550ºC, and
the interface temperature below 480ºC.
–
The liquid breeder is self-cooled; i.e., in the breeder region, the
liquid serves as breeder and coolant. The temperature of the
breeder can be kept higher than the structure temperature
15
through design, leading to higher thermal efficiency.
EU – The Helium-Cooled Lead Lithium (HCLL)
DEMO Blanket Concept
Module box
(container & surface
heat flux extraction)
Breeder cooling
unit (heat extraction
from PbLi)
[18-54]
mm/s
[0.5-1.5]
mm/s
Stiffening structure
(resistance to accidental in-box
pressurization i.e He leakage)
He collector system
(back)
HCLL PbLi flow scheme
16
Li/Vanadium Blanket Concept
Vanadium structure
Li
Lithium
Secondary Shield
Li
Primary Shield
Li
Reflector
Breeding Zone
(Li flow)
Primary shield
(Tenelon)
Secondary shield
(B4C)
Reflector
Vanadium structure
Lithium
17
Pathway Toward Higher Temperature Through Innovative
Designs with Current Structural Material (Ferritic Steel):
Dual Coolant Lead-Lithium (DCLL) FW/Blanket Concept
 First wall and ferritic steel structure
cooled with helium
 Breeding zone is self-cooled
 Structure and Breeding zone are
separated by SiCf/SiC composite
flow channel inserts (FCIs) that
 Provide thermal insulation to
decouple PbLi bulk flow
temperature from ferritic steel
wall
 Provide electrical insulation to
reduce MHD pressure drop in
the flowing breeding zone
DCLL Typical Unit Cell
Pb-17Li exit temperature can be significantly higher than the
operating temperature of the steel structure  High Efficiency
18
US DCLL DEMO Blanket Module
19
Flows of electrically conducting coolants will
experience complicated
magnetohydrodynamic (MHD) effects
What is magnetohydrodynamics (MHD)?
– Motion of a conductor in a magnetic field produces an EMF that can
induce current in the liquid. This must be added to Ohm’s law:
j   (E  V  B )
– Any induced current in the liquid results in an additional body force
in the liquid that usually opposes the motion. This body force must
be included in the Navier-Stokes equation of motion:
V
1
1
 (V  )V   p   2 V  g  j  B
t


– For liquid metal coolant, this body force can have dramatic impact
on the flow: e.g. enormous MHD drag, highly distorted velocity
profiles, non-uniform flow distribution, modified or suppressed
turbulent fluctuations
20
Main Issue for Flowing Liquid Metal in Blankets:
MHD Pressure Drop
Feasibility issue – Lorentz force resulting from LM
motion across the magnetic field generates MHD
retarding force that is very high for electrically
conducting ducts and complex
geometry flow elements
Thin wall MHD pressure drop formula
p MHD  LJB  LVB
p, pressure
L, flow length
J, current density
B, magnetic induction
V, velocity
, conductivity (LM or wall)
a,t, duct size, wall thickness
2  wt w
a


c
21
Inboard is the critical limiting
region for LM blankets
– B is very high! 10-12T
– L is fixed to reactor height by
poor access
– a is fixed by allowable
shielding size
– Tmax is fixed by material limits
Combining Power balance formula
m  PNW LLa c p T
L
With Pipe wall stress formula
S  pa / t w
With thin wall MHD pressure drop formula
(previous slide) gives:
PNWL L2 B 2 w
S
ac p T
(Sze, 1992)
Pipe stress is INDEPENDENT of wall thickness to first order
and highly constrained by reactor size and power!
22
No pipe stress window for inboard blanket operation for
Self-Cooled LM blankets (e.g. bare wall Li/V)
(even with aggressive assumptions)
– 3D features like flow
distribution and collection
manifolds
– First wall cooling likely
requiring V ~ 1 m/s
300
Unacceptable
250
Pipe Wall Stress, S (MPa)
 Pipe stress >200 MPa will
result just to remove nuclear
heat
 Higher stress values will result
when one considers the real
effects of:
U ~ 0.16 m/s
Pmax ~ 5-10 MPa
200
Marginal
150
100
Allowable
50
ARIES-RS
ITER
0
1
2
3
4
5
6
7
8
9
10
11
12
13
14
Inboard Field, B (T)
Best Possible DEMO Base Case
for bare wall Li/V:
NWL = 2.5 MW/m2
L = 8 m, a = 20 cm
T = 300K
23
15
Large MHD drag results in large MHD
pressure drop
Conducting walls
Insulated walls
Lines of current enter the low
resistance wall – leads to very
high induced current and high
pressure drop
1
0.8
0.6
0.4
1
0.8
0.6
0.4
0.2
0.2
0
0
-0.2
-0.2
All current must close in the
liquid near the wall – net drag
from jxB force is zero
-0.4
-0.6
-0.8
-1
•
•
-0.6
-0.8
-1
-1
-1
•
-0.4
-0.8
-0.6
-0.4
-0.2
0
0.2
0.4
0.6
0.8
-0.8
-0.6
-0.4
-0.2
0
0.2
0.4
0.6
0.8
1
1
Net JxB body force p = cVB2
where c = (tw w)/(a )
For high magnetic field and high
speed (self-cooled LM concepts
in inboard region) the pressure
drop is large
The resulting stresses on the
wall exceed the allowable stress
for candidate structural
materials
•
•
Perfect insulators make the net
MHD body force zero
But insulator coating crack
tolerance is very low (~10-7).
–
•
It appears impossible to develop
practical insulators under fusion
environment conditions with large
temperature, stress, and radiation
gradients
Self-healing coatings have been
proposed but none has yet been
found (research is on-going) 24
Issues with the Lithium/Vanadium Concept
•
Li/V was the U.S. choice for a long time, because of its perceived simplicity.
But negative R&D results and lack of progress on serious feasibility issues
have eliminated U.S. interest in this concept as a near-term option.
Issues
•
Insulator
Insulating layer
Conducting walls
– Insulator coating is required
– Crack tolerance (10-7) appears too low to
be achievable in the fusion environment
– “Self-healing” coatings can solve the
problem, but none has yet been found
(research is ongoing)
•
Corrosion at high temperature (coupled to
coating development)
– Existing compatibility data are limited to
maximum temperature of 550ºC and do
not support the BCSS reported corrosion
limit of 5mm/year at 650ºC
Leakage current
•
•
Electric currents lines
Crack
Tritium recovery and control
Vanadium alloy development is very costly and requires a very long time to
complete
25
Proposed US DCLL TBM Cutaway
PbLi Flow
Channels
US DCLL TBM –
Cutaway Views
PbLi
SiC FCI
He-cooled
First Wall
He
484 mm
2 mm gap
He
26
Flow Channel Inserts are a critical element
of the high outlet temperature DCLL
 FCIs are roughly box channel
shapes made from some material
with low electrical and thermal
conductivity
– SiC/SiC composites and SiC foams
are primary candidate materials
 They will slip inside the He Cooled
RAFS structure, but not be rigidly
attached
 They will slip fit over each other,
but not be rigidly attached or
sealed
 FCIs may have a thin slot or holes
in one wall to allow better pressure
equalization between the PbLi in the
main flow and in the gap region
 FCIs in front channels, back channels, and access pipes will
be subjected to different thermal and pressure conditions; and
will likely have different designs and thermal and electrical
property optimization
27
Example: Interaction between MHD flow and FCI behavior
are highly coupled and require fusion environment
 PbLi flow is strongly influenced by MHD interaction with
plasma confinement field and buoyancy-driven
convection driven by spatially non-uniform
volumetric nuclear heating
 Temperature and thermal stress of
SiC FCI are determined by this MHD flow
and convective heat transport processes
 Deformation and cracking of the FCI depend on
FCI temperature and thermal stress coupled with earlylife radiation damage effects in ceramics
 Cracking and movement of the FCIs will strongly
influence MHD flow behavior by opening up new
conduction paths that change electric current profiles
Similarly, coupled phenomena in tritium
permeation, corrosion, ceramic breeder
thermomechanics, and many other
blanket and material behaviors
FCI temperature, stress
and deformation
28
Key DCLL parameters in three
blanket scenarios
Parameter
DEMO
ITER H-H ITER D-T
Surface heat flux, Mw/m2
0.55
0.3
0.3
Neutron wall load, Mw/m2
3.08
-
0.78
500/700
470/~450
360/470
0.22x0.22x2
0.066x0.12x1.6
0.066x0.12x1.6
PbLi velocity, m/s
0.06
0.1
0.1
Magnetic field, T
4
4
4
PbLi In/Out T, C
2a x 2b x L, m
Re
61,000
30,500
Ha
11,640
6350
Gr
3.521012
7.22109
*Velocity, dimensions, and dimensionless parameters are for the poloidal flow.
DEMO parameters are for the outboard blanket.
Abdou Lecture 4
MHD / Heat Transfer analysis
(DEMO)
Pb-Li
GAP
FCI
Temperature Profile for
Model DEMO Case
FS
• Parametric analysis was
performed for poloidal flows
to access FCI effectiveness
as electric/thermal insulator
kFCI = 2 W/m-K
FCI = 5 S/m
• SiC=1-500 S/m, kSiC=2-20
W/m-K
• Strong effect of SiC on the
temperature field exists via
changes in the velocity
profile
• FCI properties were
preliminary identified:
SiC~100 S/m, kSiC~2 W/m-K
20 S/m
100 S/m
30
Molten Salt Blanket Concepts
• Lithium-containing molten salts are used as the coolant
for the Molten Salt Reactor Experiment (MSRE)
• Examples of molten salt are:
– Flibe: (LiF)n · (BeF2)
– Flinabe: (LiF-BeF2-NaF)
• The melting point for flibe is high (460ºC for n = 2, 380ºC
for n = 1)
• Flinabe has a lower melting point (recent measurement
at SNL gives about 300ºC)
• Flibe has low electrical conductivity, low thermal
conductivity
31
Molten Salt Concepts: Advantages and Issues
Advantages
•
Very low pressure operation
•
Very low tritium solubility
•
Low MHD interaction
•
Relatively inert with air and water
•
Pure material compatible with many structural materials
•
Relatively low thermal conductivity allows dual coolant concept (high thermal
efficiency) without the use of flow-channel inserts
Disadvantages
•
High melting temperature
•
Need additional Be for tritium breeding
•
Transmutation products may cause high corrosion
•
Low tritium solubility means high tritium partial pressure (tritium control
problem)
•
Limited heat removal capability, unless operating at high Re (not an issue for
32
dual-coolant concepts)
Summary of MHD issues of liquid
blankets
MHD issue
Self-cooled LM
blanket
Dual-coolant
LM blanket
He-cooled LM
blanket
MS blanket
MHD pressure
drop and flow
distribution
Very important
Important
Inlet manifold ?
Not important
Electrical
insulation
Critical issue
Required for IB
blanket
Not needed ?
Not needed
MHD
turbulence
Cannot be
ignored
Important.
2-D turbulence
Not applicable
Critical issue
(self-cooled)
Buoyancy
effects
Cannot be
ignored
Important
May effect
tritium transport
Have not been
addressed
MHD effects
on heat
transfer
Cannot be
ignored
Important
Not important
Very important
(self-cooled)
Abdou Lecture 4
Issues and R&D on Liquid Metal Breeder
Blankets
• Fabrication techniques for SiC Inserts
• MHD and thermalhydraulic experiments on SiC
flow channel inserts with Pb-Li alloy
• Pb-Li and Helium loop technology and out-ofpile test facilities
• MHD-Computational Fluid Dynamics simulation
• Tritium permeation barriers
• Corrosion experiments
• Test modules design, fabrication with RAFS,
preliminary testing
• Instrumentation for nuclear environment
34
Fusion environment is unique and complex:
multi-component fields with gradients

Neutrons (fluence, spectrum, temporal and

spatial gradients)
• Radiation Effects (at relevant temperatures, 
stresses, loading conditions)
• Bulk Heating
• Tritium Production
• Activation

Heat Sources (magnitude, gradient)
•
•

Bulk (from neutrons and gammas)
Surface
Synergistic Effects
•
•

Particle Flux (energy and density,
gradients)
Magnetic Field (3-component with
gradients)
• Steady Field
• Time-Varying Field
Mechanical Forces
• Normal/Off-Normal

Thermal/Chemical/Mechanical/
Electrical/Magnetic Interactions
Combined environmental loading conditions
Interactions among physical elements of components
Multi-function blanket in multi-component field environment leads to:
- Multi-Physics, Multi-Scale Phenomena
Rich Science to Study
- Synergistic effects that cannot be anticipated from simulations & separate effects
tests. Even some key separate effects in the blanket can not be produced in non-fusion
facilities (e.g. volumetric heating with gradients)
A true fusion environment is ESSENTIAL to Activate mechanisms that
cause prototypical coupled phenomena and integrated behavior
35
ITER Provides Substantial Hardware Capabilities
for Testing of Blanket System
TBM System (TBM + T-Extrac, Bio-shield
Heat Transport/Exchange…)
A PbLi loop
Transporter located in
the Port Cell Area
2.2 m
ITER has allocated 3 ITER
equatorial ports (1.75 x 2.2
m2) for TBM testing
Each port can accommodate
only 2 Modules (i.e. 6 TBMs
He pipes to
TCWS
Equatorial Port
Plug Assy.
Vacuum Vessel
max)
But, 12 modules are
proposed by the parties
TBM
Assy
Port
Frame
Aggressive competition for Space
36
Challenging
Fusion Nuclear Technology Issues
1. Tritium Supply &
Tritium Self-Sufficiency
2. High Power Density
3. High Temperature
4. MHD for Liquid Breeders / Coolants
5. Tritium Control (Permeation)
6. Reliability / Maintainability / Availability
7. Testing in Fusion Facilities
37
Tritium Self-Sufficiency
• TBR ≡ Tritium Breeding Ratio = N  / N 
N  = Rate of tritium production (primarily in the blanket)
N  = Rate of tritium consumption (burnt in plasma)
Tritium self-sufficiency condition: Λa > Λr
Λr = Required tritium breeding ratio
Λr is 1 + G, where G is the margin required to: a) compensate for losses and
radioactive decay between production and use, b) supply inventory for start-up of
other fusion systems, and c) provide a hold-up inventory, which accounts for the time
delay between production and use as well as reserve storage. Λr is dependent on
many system parameters and features such as plasma edge recycling, tritium
fractional burn up in the plasma, tritium inventories, doubling time,
efficiency/capacity/reliability of the tritium processing system, etc.
Λa = Achievable breeding ratio
Λa is a function of FW thickness, amount of structure in the blanket, presence of
stabilizing shell materials, PFC coating/tile/materials, material and geometry for
divertor, plasma heating, fueling and penetration.
38
Current physics and technology concepts lead to a
“narrow window” for attaining Tritium self-sufficiency
Required TBR
td = doubling time
td=1 yr
td=5 yr
Fusion power
1.5GW
Reserve time
2 days
Waste removal efficiency 0.9
(See paper for details)
Max achievable
TBR ≤ 1.15
td=10 yr
“Window” for
Tritium self
sufficiency
Fractional burn-up [%]
39
Fuel Cycle Dynamics
The D-T fuel cycle includes many components whose operation parameters and
their uncertainties impact the required TBR
Fueling
Plasma
Fuel management
Plasma exhaust
processing
Impurity
separation
FW coolant
processing
Plasma
Facing
Component
Solid waste
Breeder Blanket
Fuel inline
storage
Impurity processing
Coolant
tritium
recovery
system
PFC
Coolant
Blanket
Coolant
processing
Tritium
waste
treatment
(TWT)
Tritium
shipment/permanent
storage
•ß: Tritium fraction
burn-up
Isotope
separation
system
•Ti: mean T
residence time in
each component
•Tritium inventory
in each component
Water stream
and air
processing
waste
Blanket tritium
recovery system
Only for solid breeder or liquid
breeder design using separate
coolant
Examples of key
parameters:
Only for liquid breeder
as coolant design
•Doubling time
•Days of tritium
reserves
•Extraction
inefficiency in
plasma exhaust
processing
40
41
Structural Materials
• Key issues include thermal stress capacity, coolant compatibility,
waste disposal, and radiation damage effects
• The 3 leading candidates are ferritic/martensitic steel, V alloys and
SiC/SiC (based on safety, waste disposal, and performance
considerations)
• The ferritic/martensitic steel is the reference
structural material for DEMO
(Commercial alloys (Ti alloys, Ni base superalloys, refractory alloys, etc.) have been
shown to be unacceptable for fusion for various technical reasons)
Structural
Materia l
Coolant/Tritium Breeding Material
Li/Li
He/PbLi
H2O/PbLi
He/Li ceramic H2O/Li ceramic FLiBe/FLiBe
Ferritic steel
V alloy
SiC/SiC
42
43
44
Fission (PWR)
Fusion structure
Coal
Tritium in fusion
45
Lessons learned:
The most challenging problems in FNT
are at the INTERFACES
• Examples:
– MHD insulators
– Thermal insulators
– Corrosion (liquid/structure interface temperature limit)
– Tritium permeation
• Research on these interfaces must be done
jointly by blanket and materials researchers
46
Need for High Power Density Capability
A. To improve potential attractiveness of fusion power
compared to other energy sources (e.g., fission)
PWR
Average core power
density (MW/m3)
96
BWR LMFBR ITER-Type
56
240
0.4
B. Some plasma confinement schemes have the
potential to produce burning plasmas with high
power density
– FW/Blanket/Divertor concepts developed in the 1970s and ’80s have
limitations on power density capability (wall load and surface heat flux)
– Substantial PROGRESS has been made in this area over the past
several years (e.g. in the APEX Study in the US )
47
Reliability/Maintainability/Availability is one of the remaining
“Grand Challenges” to Fusion Energy Development. Chamber
Technology R&D is necessary to meet this Grand Challenge.
Need High Power Density/Physics-Technology Partnership
-High-Performance Plasma
-Chamber Technology Capabilities
Need Low
Failure Rate
C  i  replacement cost  O & M
COE 
P fusion  Availability  M  h th
Energy
Multiplication
Need High Temp.
Energy Extraction
Need High Availability / Simpler Technological and Material Constraints
(1 / failure rate )
1 / failure rate  replacement time
 Need Low Failure Rate:
- Innovative Chamber Technology
 Need Short Maintenance Time:
- Simple Configuration Confinement
- Easier to Maintain Chamber Technology
48
MTBF
Availability =
MTBF + MTTR
Current plasma confinement schemes and
configurations have:
– Relatively long MTTR (weeks to months)
Required MTBF must be high
– Large first wall area
Unit failure rate must be very low
MTBF ~ 1/(area • unit failure rate)
Reliability requirements are more demanding
than for other non-fusion technologies
49
ed
(R
)
600
5
400
200
0
Expected
0
1
2
C
A
0
3
MTBF per Blanket Segment(FPY)
800
N
ee
d
MTBF per Blanket System(FPY)
10
MTTR (Months)
The reliability requirements on the Blanket/FW (in current
confinement concepts that have long MTTR > 1 week) are
most challenging and pose critical concerns. These must be
seriously addressed as an integral part of the R&D pathway to
DEMO. Impact on ITER is predicted to be serious. It is a DRIVER
for CTF.
50
“Reliability Growth”
Upper statistical confidence level as a function of test time in
multiples of MTBF for time terminated reliability tests (Poisson
distribution).
Results are given for different numbers of failures.
1.0
Number of Failures 0
Confidence Level
0.8
Example,
TYPICAL
TEST
SCENARIO
1
To get 80% confidence
in achieving a particular
value for MTBF, the
total test time needed
is about 3 MTBF (for
case with only one
failure occurring during
the test).
2
0.6
3
0.4
4
0.2
0.0
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
4.0
4.5
5.0
Test Time in Multiplies of Mean-Time-Betw een-Failure (MTBF)
Reference: M. Abdou et. al., "FINESSE A Study of the Issues, Experiments and Facilities for Fusion Nuclear
Technology Research & Development, Chapter 15 (Figure 15.2-2.) Reliability Development Testing Impact on
Fusion Reactor Availability", Interim Report, Vol. IV, PPG-821, UCLA,1984. It originated from A. Coppola,
"Bayesian Reliability Tests are Practical", RADC-TR-81-106, July 1981.
51
Summary of Critical R&D Issues for Fusion Nuclear Technology
1.
D-T fuel cycle tritium self-sufficiency in a practical system
depends on many physics and engineering parameters / details: e.g. fractional burn-up
in plasma, tritium inventories, FW thickness, penetrations, passive coils, etc.
2. Tritium extraction and inventory in the solid/liquid breeders
under actual operating conditions
3. Thermomechanical loadings and response of blanket and PFC
components under normal and off-normal operation
4. Materials interactions and compatibility
5. Identification and characterization of failure modes, effects, and
rates in blankets and PFC’s
6. Engineering feasibility and reliability of electric (MHD) insulators
and tritium permeation barriers under thermal / mechanical /
electrical / magnetic / nuclear loadings with high temperature and
stress gradients
7. Tritium permeation, control and inventory in blanket and PFC
8. Lifetime of blanket, PFC, and other FNT components
9. Remote maintenance with acceptable machine shutdown time 52
ITER TBM is the Necessary First Step to enable
future Engineering Development
Role of ITER TBM
Component Engineering
Development &
Reliability Growth
Fusion “Break-in” &
Scientific Exploration
Engineering Feasibility
& Performance
Verification
Stage I
Stage II
Stage III
~ 0.3 MW-y/m2
1 - 3 MW-y/m2
> 2 - 4 MW-y/m2
Sub-Modules/
Modules Size Tests
Module Size
Tests
Module / Sectors
Size Tests
• Initial exploration of coupled
phenomena in a fusion environment
• Uncover unexpected synergistic effects,
Calibrate non-fusion tests
• Impact of rapid property changes in
early life
• Integrated environmental data for
model improvement and simulation
benchmarking
• Develop experimental techniques and
test instrumentation
• Screen and narrow the many material
combinations, design choices, and
blanket design concepts
• Uncover unexpected synergistic
effects coupled to radiation
interactions in materials, interfaces,
and configurations
• Verify performance beyond beginning
of life and until changes in properties
become small (changes are substantial
2
up to ~ 1-2 MW · y/m )
• Initial data on failure modes & effects
• Establish engineering feasibility of
blankets (satisfy basic functions &
performance, up to 10 to 20 % of
lifetime)
• Select 2 or 3 concepts for further
development
D
E
M
O
• Identify lifetime limiting failure modes
and effects based on full environment
coupled interactions
• Failure rate data: Develop a data base
sufficient to predict mean-timebetween-failure with confidence
• Iterative design / test / fail / analyze /
improve programs aimed at reliability
growth and safety
• Obtain data to predict mean-time-toreplace (MTTR) for both planned
outage and random failure
• Develop a data base to predict overall
availability of FNT components in
DEMO
53
Tritium breeding capabilities are needed for the continued operation
of successful ITER and fusion development
Production & Cost:
CANDU Reactors: 27 kg from over 40 years,
$30M/kg (current)
Fission reactors: 2–3 kg/year/reactor,
$84M-$130M/kg (per DOE Inspector General*)
*www.ig.energy.gov/documents/CalendarYear2003/ig-0632.pdf
 A Successful ITER will exhaust most of
the world supply of tritium
 ITER extended performance phase and
any future long pulse burning plasma will
need tritium breeding technology
 TBMs are critical to establishing the
knowledge base needed to develop even
this first generation of breeding capability
Projected Ontario (OPG) Tritium Inventory (kg)
Tritium Consumption in Fusion is HUGE! Unprecedented!
55.8 kg per 1000 MW fusion power per year
30
CANDU
Supply
w/o Fusion
25
20
15
10
5
0
1995 2000 2005
1000 MW Fusion
10% Avail,
TBR 0.0
ITER-FEAT
(2004 start)
2010 2015 2020 2025 2030 2035 2040 2045
Year
TBM helps solve the Tritium Supply Issue for fusion development
(at a fraction of the cost of purchasing tritium from fission reactor sources!) 54
BACKUP SLIDES
55
Will the development of high-temperature structural
material lead to more attractive blankets?
Not necessarily (unless we can solve the interface problems)
1. Vanadium alloys (high temperature capability)
• V is compatible only with liquid lithium
• Flowing liquid Li requires MHD insulators
• Tolerable crack fraction is estimated to be very low (< 10-7)—much lower
than can be achieved with real coatings
• “Self-healing” coating R&D results are negative for non-isothermic systems
• Laminated layer insulators (alternating layers of insulator and metallic
protection layer) were proposed, but V layer needs to be too thin
2. High-temperature advanced ferritic steels (ODS, NFS)
• May potentially operate up to ~700ºC (compared to 550ºC for EUROFER
and F82H)
• At present, we cannot utilize such advanced high-temperature ferritics
• Li and LiPb interface temperature (Tint) is limited by corrosion to ~500ºC
Unless corrosion temperature limit is improved, EUROFER and F82H
are satisfactory
• Solid breeder/structure interface cannot be increased much above 400–
500ºC (to have adequate temperature window for T-release and TBR)
56
Tritium Control and Management
• Tritium control and management will be one of the most difficult
issues for fusion energy development, both from the technical
challenge and from the “public acceptance” points of view.
• Experts believe the T-control problem is underestimated
(maybe even for ITER!)
• The T-control problem in perspective:
– The scale-up from present CANDU experience to ITER and DEMO
is striking:
The quantity of tritium to be managed in the ITER fuel cycle is much
larger than the quantities typically managed in CANDU (which represents
the present-day state of practical knowledge).
– The scale-up from ITER to DEMO is orders of magnitude:
The amount of tritium to be managed in a DEMO blanket (production rate
~400 g/day) is several orders of magnitude larger than that expected in
ITER, while the allowable T-releases could be comparable.
For more details, see:
– W. Farabolini et al, “Tritium Control Modelling in an He-cooled PbLi Blanket…” paper in ISFNT-7 (this conference)
– Papers and IEA Reports by Sze, Giancarli, Tanaka, Konys, etc.
57
Why is Tritium Permeation a Problem?
• Most fusion blankets have high tritium partial
pressure:
LiPb = 0.014 Pa
Flibe = 380 Pa
He purge gas in solid breeders = 0.6 Pa
• The temperature of the blanket is high (500–700ºC)
• Surface area of heat exchanger is high, with thin walls
• Tritium is in elementary form
These are perfect conditions for tritium permeation.
• The allowable tritium loss rate is very low
(~10 Ci/day), requiring a partial pressure of ~10-9 Pa.
Challenging!
• Even a tritium permeation barrier with a permeation
reduction factor (PRF) of 100 may be still too far from
solving this problem!
58
Pillars of a Fusion Energy System
1. Confined and Controlled
Burning Plasma (feasibility)
2. Tritium Fuel Self-Sufficiency
(feasibility)
3. Efficient Heat Extraction and
Conversion (attractiveness)
4. Safe and Environmentally
Advantageous
(feasibility/attractiveness)
The Blanket is THE
KEY component
5. Reliable System Operation
(attractiveness)
Yet, No fusion blanket has ever been built or tested!
59
Structure of TBM collaboration*
 Each 1/2 of a port is dedicated to testing of one TB concept design (one
module or several connected sub-modules).
 Each Concept design is tested by a partnership of a TB concept leader +
supporting partners
 One of the two TB leaders occupying the same port must play a role of the
Port master – responsible for integration of the given port
 Port Master has main responsibility of integration of 2 concepts in the
same port frame and preparation of the integrated testing program
(replacement strategy) for the port
ITER has only 3 ports  Only 6 TBMs
can be tested
List of TBM Design Proposals for day-one
 Helium-cooled Lithium-Lead TBM (2
designs)
 Dual-coolant (He+Lithium-Lead) TBM (2
designs)
 He-cooled Ceramic Breeder/Beryllium
multiplier TBM (4 designs)
 Water-cooled Ceramic Breeder/Beryllium
multiplier TBM (1 design)
 He-cooled Liquid Lithium TBM (1 design)
Self-cooled Liquid Lithium TBM (1 design)
* Initial result of TBWG DH meeting, 18-19 July 2006
1248
750
(DDDs completed by Parties)
524
1700
Horizontal
Vertical
Each Port can hold two vertically or horizontally
oriented half-port size integrated TBMs
60
US advanced Ideas to Solve the TBM testing-space
problem in ITER, while improving effectiveness of the tests
KO Submodule
JA Submodule
US Submodule
 US experiments can
focus on niche areas of
US interest and
expertise,
 While benefiting from
world resources and
testing results
 Different sub-modules
can be tailored to
address the many
different:
The back plate coolant supply
and collection manifold assembly
incorporates all connections to main
support systems. A “Lead Party” takes responsibility
for back plate and sub-module integration
– design configurations
– material options
– operating conditions
(such as flow rates,
temperatures, stresses)
– diagnostics and
experimental focus
61
Projected Ontario (OPG) Tritium
Inventory (kg)
Projections for World Tritium Supply Available to Fusion
Reveal Serious Problems
30
25
CANDU Supply
20
w/o Fusion
15
World Max. tritium supply is 27 kg
10
Tritium decays at a rate of 5.47% per year
5
0
2000
2010
2020
2030
2040
Year
Tritium Consumption in Fusion is HUGE! Unprecedented!
55.8 kg per 1000 MW fusion power per year
Production & Cost:
CANDU Reactors: 27 kg from over 40 years, $30M/kg (current)
Fission reactors: 2–3 kg per year.
It takes tens of fission reactors to supply one fusion reactor.
$84M-$130M per kg, per DOE Inspector General*
*DOE Inspector General’s Audit Report, “Modernization of Tritium Requirements Systems”, Report DOE/IG-0632, December 2003,
available at www.ig.doe.gov/pdf/ig-0632.pdf
62
Tritium Consumption in ITER
 Here is from a summary of the final design report.
Link is:
http://fusion.gat.com/iter/iter-ga/images/pdfs/cost_estimates.pdf
 9.4.3 Fuel Costs
The ITER plant must be operated, taking into account the available
tritium externally supplied. The net tritium consumption is 0.4
g/plasma pulse at 500 MW burn with a flat top of 400 s
 “The total tritium received on site during the first 10 years of
operation, amounts to 6.7 kg.”
 “whereas the total consumption of tritium during the plant life time
may be up to 16 kg to provide a fluence of 0.3 MWa/m2 in average
on the first wall”
 “This corresponds, due to tritium decay, to a purchase of about 17.5
kg of tritium. This will be well within, for instance, the available
Canadian reserves.”
63
It is important to precisely understand the state-of-theart in blanket and material research, and the role of ITER
 Over the past 30 years the fusion nuclear technology and
materials programs have spent much effort on developing
theories and models of phenomena and behavior.
– But, these are for idealized conditions based on understanding of
single effects. There has never been a single experiment in a fusion
environment!
Are they scientifically valid?
 ITER will provide the 1st opportunity to test these theories and
models in a real fusion environment.
Only the Parties who will do
successful, effective TBM experiments in ITER will have
the experimentally-validated scientific basis to embark on
the engineering development of tritium breeding blankets
64
Another Look at the DCLL Unit Cell
65
A Simplified DCLL DEMO System
450C He
350C He
From/To Helium Loops and Brayton
Cycle Power Conversion System
450C 650C
Blanket
Module
Tritium Extraction
Heat Exchanger
Coaxial Feed Pipes
• PbLi Hot leg flows in
inner pipe (700C)
• PbLi Cold leg flows in
outer annulus (450C)
• Cold leg cools Pipe walls
and TX/HX shells
From/To Tritium
Processing System
Pump
Cold
Trap,
Chem.
Control
66
Div He Tout
Coolant Routing Through HX Coupling
Blanket and Divertor to Brayton Cycle
T
Typical Fluid Temperatures in HX
for DCLL ARIES-CS
Blkt LiPb
452°C
Blkt He
369°C
He
from
Div He
Divertor
Tin
(Pth,fus+Pfrict)Div,He
He THX,out
LiPb
Tout
Blkt LiPb (711°C)
+ Div He (711°C)
580°C
~681°C
441°C
Cycle He
Pb-17Li
from
Blanket
349°C
ZHX
Power Parameters for DCLL in ARIES-CS
Fusion Thermal Power in Reactor Core
2650 MW
Fusion Thermal Power in Pb-17Li
1420 MW
Fusion Thermal Power in Blkt He
1030 MW
Friction Thermal Power in Blkt He
119 MW
Fusion Thermal Power in Div He
201 MW
Friction Thermal Power in Div He
29 MW
Total Power
2790 MW
Overall Brayton Cycle Efficiency
0.42
(Pth,fus)Blkt,LiPb
LiPb Tin
Brayton
Cycle
Blkt He Tout
He
from
Blanket
(Pth,fus+Pfrict)Blkt,He
Blkt He Tin
He THX,in
67
Blanket systems are complex and have many
integrated functions, materials, and interfaces
[18-54] mm/s
[0.5-1.5] mm/s
Neutron Multiplier
Be, Be12Ti (<2mm)
Tritium Breeder
Li2TiO3 (<2mm)
PbLi flow
scheme
First Wall
(RAFS, F82H)
Surface Heat Flux
Neutron Wall Load
68
TBM is an integral part of
ITER Schedule, Safety, and Licensing
ITER Schedule
1st 10 yrs
Early H-H Phase TBM Testing is mandated by ITER IO and licensing team:
1. Optimize plasma control in the presence of Ferritic Steel TBMs
2. Qualify port integration and remote handling procedures
3. License ITER with experimental TBMs for D-T operation
69
Shield
Blanket
Vacuum vessel
Radiation
Plasma
Neutrons
First Wall
Tritium breeding zone
Coolant for energy
conversion
Magnets
70
Helium-Cooled Pebble Breeder Concept for EU
Helium-cooled stiffening grid
Breeder unit
71
FW channel
US DCLL Concept, 4
Summary of MHD pressure drops for ITER TBM
Flow
Pi, MPa
Pi/P
1. Front channel
0.38410-3
0.13
2. Return channel
0.48510-3
0.16
3. Concentric pipe
(internal, uniform B-field)
15.410-3
5.1
4. Concentric pipe
(annulus, uniform B-field)
0.0286
9.5
5. Concentric pipe
(internal, fringing B-field)
0.0585
19.3
6. Concentric pipe
(annulus, fringing B-field)
0.0585
19.3
7. Inlet manifold
0.070-0.140
23.2
8. Outlet manifold
0.070-0.140
23.3
Total
0.302-0.442
100
1
2
8
4
3
7
72
FW Heat Flux and Neutron Wall Load
Simplified DCLL Blanket Module Flow Scheme
PbLi
(625C)
PbLi Out
(700C)
 All structural walls are RAFS actively cooled
by He
 Cold PbLi flows up the FW (where volumetric
heating is strongest), turns, and flows back
down the back of the blanket module
 SiC FCIs separates and insulates the flowing
PbLi from the RAFS walls
 FCIs are loosely slip-fit together, and GAPs
between FCIs and structure is filled in by
nearly stagnant PbLi
SiC FCIs
PbLi In (450C)
Helium-cooled RAFS
FW and structure
Gap between FCI and Structure
(Filled with nearly stagnant PbLi)
 The interface temperature between the
RAFS structure and gap PbLi is
controlled by the He cooling, and
kept < 500C.
73
Idea of Coaxial Pipe for PbLi feed lines
– use FCI to insulate hot leg from cold
Coaxial Pipe Outer Wall
Outer FCI (For MHD insulation)
To Reactor ~450C
From Reactor ~700C
Coaxial Pipe Inner Wall (~500C)
PbLi Gap (~500C)
Inner FCI
• Inner FCI insulates inner
hot leg PbLi flow
• Allows outer cold leg PbLi
flow to cool Inner pipe wall
and PbLi gap to < 500C
• Same principle can be applied for TX and HX outer
shells
• Allows use of ordinary RAFS for almost all
structure
74
DCLL should be effective in reducing
MHD pressure drop to manageable levels
 Higher outlet temperature due
to FCI thermal insulation
allows large coolant delta T
in breeder zone, resulting in
lower mass flow rate
requirements and thus lower
velocity.
 Electrical insulation provided
by insert reduces bare wall
pressure drop by a factor of
10-100.
500
(dP/dx)0 / (dP/dx)
 Low velocity due to
elimination of the need for
FW cooling reduces MHD
pressure drop.
PEH
PES
400
300
200
100
0
1
10
100
1000
Electrical conductivity, S/m
75
ITER Objectives
Programmatic
• Demonstrate the scientific and technological feasibility of fusion
energy for peaceful purposes.
Technical
• Demonstrate extended burn of DT plasmas, with steady state as the
ultimate goal.
• Integrate and test all essential fusion power reactor technologies
and components.
• Demonstrate safety and environmental acceptability of fusion.
76
FNT: Components from Edge of Plasma to TFC
Blanket / Divertor immediately circumscribe the plasma
Poloidal Ring Coil
Cryostat
Coil Gap
Rib Panel
Blanket
Maint.
Port
Plasma
Vacuum
Vessel
Center Solenoid Coil
Toroidal Coil
77
JAEA DEMO Design
What can be done about
MHD pressure drop?
P  cL lVB
2
c represents a measure of relative
conductance of induced current closure paths
 Lower C
 Break electrical coupling to thick
– Insulator coatings

– Flow channel inserts  load bearing channel walls
– Elongated channels with anchor links  Force long current path
or other design solutions

 Lower V
– Heat transfer enhancement or separate coolant to lower velocity
required for first wall/breeder zone cooling
– High temperature difference operation to lower mass flow
 Lower B,L
– Outboard blanket only (ST)
 Lower  (molten salt)
78
Blanket Materials
1.
Tritium Breeding Material (Lithium in some form)
Liquid: Li, LiPb (83Pb 17Li), lithium-containing molten salts
Solid: Li2O, Li4SiO4, Li2TiO3, Li2ZrO3
2.
Neutron Multiplier (for most blanket concepts)
Beryllium (Be, Be12Ti)
Lead (in LiPb)
3.
Coolant
– Li, LiPb
4.
– Molten Salt
– Helium
– Water
Structural Material
–
Ferritic Steel (accepted worldwide as the reference for DEMO)
–
Long-term: Vanadium alloy (compatible only with Li), and SiC/SiC
5.
MHD insulators (for concepts with self-cooled liquid metals)
6.
Thermal insulators (only in some concepts with dual coolants)
7.
Tritium Permeation Barriers (in some concepts)
8.
Neutron Attenuators and Reflectors
79
Neutron Multipliers
Examples of Neutron Multipliers
Beryllium, Lead
• Almost all concepts need a
neutron multiplier to achieve
adequate tritium breeding.
(Possible exceptions: concepts with
Li and Li2O)
Be-9 (n,2n) and Pb(n,2n)
Cross-Sections- JENDL-3.2 Data
10
• Desired characteristics:
– Large (n, 2n) cross-section with
low threshold
– Small absorption cross-sections
1
Be-9 (n,2n)
Pb (n,2n)
• Candidates:
– Beryllium is the best (large n, 2n
with low threshold, low
absorption)
– Be12Ti may have the advantage
of less tritium retention, less
reactivity with steam
– Pb is less effective except in
LiPb
– Beryllium results in large energy
multiplication, but resources are
limited
0.1
9Be
(n,2n)
Pb (n,2n)
0.01
0.001
10
6
10
Neutron Energy (eV)
7
80