Overview of Fusion Technology
Mohamed Abdou
Distinguished Professor, Mechanical and Aerospace Engineering Department
Director, Center for Energy Science and Technology Advanced Research (CESTAR)
Director, Fusion Science and Technology Center
University of California Los Angeles (UCLA)
Seminar at Xi'an Jiaotong University
May 2006
Outline
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World energy needs
Fusion: Energy source for the XXI century
ITER
ITER Test Blanket
Principles of Fusion Nuclear Technology
Blanket concepts
He-Cooled Ceramic Breeder Blanket (HCBB)
Dual Coolant Lead-Lithium Concept (DCLL)
SiC/SiC Flow Channel Insert (FCI)
Liquid Walls
MHD Code Development
Summary
What is Nuclear Fusion?
•
•
Nuclear Fusion is the energy-producing process taking place in the core of
the Sun and stars
The core temperature of the Sun is about 15 million °C. At these
temperatures hydrogen nuclei fuse to give Helium and Energy. The
energy sustains life on Earth via sunlight
Fusion Reactions
• Deuterium – from water
(0.02% of all hydrogen is heavy hydrogen or
deuterium)
• Tritium – from lithium
(a light metal common in the Earth’s crust)
Deuterium + Lithium → Helium + Energy
This fusion cycle (which has the fastest
reaction rate) is of interest for Energy
Production
Figure 1.1 Energy use (in gigajoules) vs. GDP (on a purchasing power parity basis) for selected countries on a per capita basis.
Data from the International Energy Agency. Upper line indicates ratio for the US; lower line indicates ratio for Japan and several
Western European countries.
Figure 1.2. Human development index vs. per capita electricity use for selected countries. Taken from S. Benka, Physics
Today (April 2002), pg 39, and adapted from A. Pasternak, Lawrence Livermore National Laboratory rep. no. UCRL-ID140773.
The World, particularly in developing
countries, needs a New Energy Source
• Growth in world population and growth in energy demand from
increased industrialisation/affluence will lead to an Energy Gap which will
be increasingly difficult to fill with fossil fuels
• Without improvements in efficiency we will need 80% more energy by 2020
• Even with efficiency improvements at the limit of technology we would still
need 40% more energy
World Energy Scene* (I)
1) The world uses a lot of energy
Average power consumption = 13.6 TWs, or 2.2 kWs per person
World energy market ~ $3 trillion/yr (electricity ~$1 trillion/yr)
- very unevenly (OECD 6.2 kW/person; Bangladesh
0.20 kW/person; China 1.3kW/person)
2) World energy use is expected to grow
- growth necessary to lift billions of people out of poverty
3) 80% is generated by burning fossil fuels
 climate change & debilitating pollution
- which won’t last for ever
Need major new (clean) energy sources
- requires new technology
*See Sir Chris Llewellyn-Smith, FPA, October 11, 2005
Future Energy Use
 The International Energy Agency (IEA) expects the
world’s energy use to increase 60% by 2030 (while
population expected to grow from 6.2B to 8.1B) driven largely by growth of energy use and population
in India (current use = 0.7 kWs/person, vs. OECD
average of 6.2 kWs/person) and China (current use =
1.3 kWs/person)
 Strong link between energy use and the Human
Development Index (HDI ~ life expectancy at birth +
adult literacy and school enrolment + gross national
product per capita at purchasing power parity) – need
increased energy use to lift billions out of poverty
Carbon dioxide levels over the last 60,000
years - we are provoking the atmosphere!
Source University of Berne and National Oceanic, and Atmospheric Administration
Meeting the Energy Challenge Requires:
Fiscal measures to change the behaviour of consumers,
and provide incentives to expand use of low carbon
technologies
Actions to improve efficiency (domestic, transport, …)
Use of renewables where appropriate (although locally
useful, not hugely significant globally)
BUT only four sources capable in principle of meeting a
large fraction of the world’s energy needs:
• Burning fossil fuels (currently 80%) - develop & deploy CO2 capture
and storage
• Solar - seek breakthroughs in production and storage
• Nuclear fission - hard to avoid if we are serious about reducing fossil
fuel burning (at least until fusion available)
• Fusion - with so few options, we must develop fusion as fast as
possible, even if success is not 100% certain
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
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
ITER is a collaborative effort among Europe,
Japan, US, Russia, China, South Korea, and India
ITER Location
Caradache (France)
Rokkasho (Japan)
Cadarache was selected as the ITER construction site. There will be
some facilities in Rokassho under the “Broader Approach” agreement.
Fusion Power Station Schematic
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
Shield
Blanket
Vacuum vessel
Radiation
Plasma
Neutrons
First Wall
Tritium breeding zone
Coolant for energy
conversion
Magnets
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
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
Heat and Radiation Loads on First Wall
• Neutron Wall Load ≡ Pnw
Pnw = Fusion Neutron Power Incident on the First Wall per unit area
= JwEo
Jw = fusion neutron (uncollided) current on the first wall
Eo = Energy per fusion neutron = 14.06 MeV
• Typical Neutron Wall Load ≡ 1-5 MW/m2
At 1 MW/m2: Jw = 4.43 x 1017 n · m-2 · s-1
• Note the neutron flux at the first wall (0-14 MeV) is about
an order of magnitude higher than Jw
• Surface heat flux at the first wall
This is the plasma radiation load. It is a fraction of the α-power
qw = 0.25 Pnw · fα
where f is the fraction of the α-power reaching the first wall
(note that the balance, 1 – f, goes to the divertor)
Tritium Breeding Blankets:
Complex component submitted to very severe working conditions,
Needed in DEMO, not present in ITER
► ITER is a unique opportunity to test breeding blanket mock-ups:
Test Blanket Modules (TBMs)
► It is an ITER mission : “ITER should test tritium breeding module concepts
that would lead in a future reactor to tritium self-sufficiency, the extraction of
high grade heat and electricity production.” (ITER SWG Report to the IC)
 TBMs have to be representative of a DEMO breeding blanket, capable of
ensuring tritium-breeding self-sufficiency using high-grade coolants for
electricity production
► The ITER TBM Program is therefore a central element in the plans of all
seven ITER Parties for the development of tritium breeding and power
extraction technology.
What is the ITER TBM Program?
Integrated testing of breeding blanket and first wall components and
materials in a Fusion Environment
•
Breeding Blankets/FWs will be tested in ITER, starting on Day One, by
inserting Test Blanket Modules (TBMs) in specially designed ports.
•
Each TBM will have its own dedicated systems for tritium recovery and
processing, heat extraction, etc. Each TBM will also need new diagnostics
for the nuclear-electromagnetic environment.
•
Each ITER Party is allocated limited space for testing two TBMs. (Number
of Ports reduced to 3. Number of Parties increased to 7).
•
ITER’s construction plan includes specifications for TBMs because of
impacts on space (port, port area, hot cell, TCWS), shielding, vacuum
vessel, remote maintenance, ancillary equipment, safety, availability, etc.
•
The ITER Test Program is managed by the ITER Test Blanket Working
Group (TBWG) with participants from the ITER International Team and
representatives of the Parties. (However, this entity may change under the new
international agreement being negotiated.)
Available TBM development time is limited: ITER Schedule
calls for TBM testing from Day 1 of H-H plasma Operation
•“Day 1” H-H Phase TBM Testing in ITER is required for:
– optimization of ITER plasma control, which can only be done with the TBMs
installed because the use of ferromagnetic structure and effects on plasma-wall
interactions.
– licensing for ITER D-T operation, which requires TBM integrity under the severe
ITER abnormal loading conditions (e.g. disruption) to be demonstrated in the H-H
phase.
• Schedule from
ITER
documents
show TBM
starting from
day 1
• It is not certain
how open ITER
will be to
fielding
unproven TBMs
in the D-T
phases
TBMs Arrangement in ITER and Interfaces
► 3 ITER equatorial ports (opening of 1.75 x 2.2 m2) devoted to TBM testing
► TBMs installed within a water-cooled steel frame (thk. 20 cm), typically half-port size
T
B
M
TBMs
tests
need a
whole
TBM
system
P
O
R
T
S
TBM
Shield
plug
Frame
The TBMs first wall
is recessed 50 mm
and protected with a
Be layer
vertical
horizontal
Sample
TBM (RF)
US ITER Test Blanket
Module Concepts
DCLL Typical Unit Cell
 The Dual-Coolant Lead-Lithium (DCLL) and
the Helium Cooled Ceramic Breeder (HCCB)
concepts have been selected for ITER testing by
the US community.
 The DCLL is chosen as an innovative concept
that provides a “pathway” to higher outlet
temperature and higher efficiency while using
current generation low-activation ferritic steel
(FS) as a structural material and SiC composite
only as a non-structural insulator.
Assembly of 3 HCCB
sub-modules in 1 half-port
DCLL Perspective cutaway
 The HCCB is chosen as the most likely
candidate for near term tritium breeding
blankets, e.g. in an extended performance phase
of ITER, while providing high grade heat for
electricity production.
 Both concepts use reduced activation ferritic
steel (RAFS) as a structural material and high
pressure helium as a coolant. RAFS maximum
operating temperature (550C) dictates the
maximum helium coolant outlet temperature.
Beryllium Pebble Bed
Solid Breeder Pebble Bed
Coolant Channel
US TBM Program has estimated its costs based on the
degree of international collaboration / cost sharing
 The high cost range scenario is for an independent US DCLL TBM
and an independent HCCB TBM (similar to EU, Japan, and other
parties independently testing two full modules.
 The baseline scenario consists of (1) an independent US DCLL TBM ,
and (2) a supporting partnership with another party (Japan or EU) on
the HCCB TBM providing only a submodules (size is 1/3 of a module)
 The low cost range scenario is defined as a leading international
partnership (with one or more ITER Parties) on DCLL TBM and a
supporting partnership on the HCCB.
Low - Partnership
on both concepts
Baseline Partnership on one
concept
High - Little
Collaboration
ITER-TBM Estimated Cost (2006 k$)
$60,392
$89,676
$119,666
Est. Escalation and Contingency
$16,927
$23,271
$30,947
Total Program Cost (escalated k$ including contingency)
$77,319
$112,947
$150,613
All costs over the next 10 years up to shipping of first
TBM systems
Comparison of Total Program Cost Ranges
for the US TBM Program
(including escalation and contingency)
$30,000
Low
Total Costs (k$)
$25,000
Baseline
$20,000
High
$15,000
$10,000
$5,000
$0
FY06
FY07
FY08
FY09
FY10
FY11
FY12
FY13
FY14
FY15
Low
$1,882
$6,842
$10,423
$13,760
$14,377
$12,121
$8,775
$5,597
$2,759
$783
Baseline
$2,749
$9,995
$15,226
$20,100
$21,002
$17,707
$12,819
$8,177
$4,030
$1,144
High
$3,666
$13,328
$20,303
$26,803
$28,006
$23,611
$17,093
$10,903
$5,374
$1,525
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
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 burnup 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.
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
– 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
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
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
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)
Helium-Cooled Pebble Breeder Concept for EU
Helium-cooled stiffening grid
Breeder unit
FW channel
Stiffening plate provides the mechanical strength
to the structural box
Radial-poloidal plate
Grooves for helium
coolant
Helium
Radial-toroidal plate
Cut view
Breeder Unit for EU Helium-Cooled Pebble Bed
Concept
Mechanisms of tritium transport (for solid breeders)
Li(n, 4He)T
Breeder
pebble
(solid/gas interface where
adsorption/desorption occurs)
Mechanisms of tritium transport
1)
2)
3)
4)
5)
Intragranular diffusion
Grain boundary diffusion
Surface Adsorption/desorption
Pore diffusion
Purge flow convection
Purge gas composition:
He + 0.1% H2
Tritium release composition:
T2, HT, T2O, HTO
“Temperature Window” for Solid Breeders
• The operating temperature of the solid breeder is limited
to an acceptable “temperature window”: Tmin– Tmax
– Tmin, lower temperature limit, is based on acceptable tritium
transport characteristics (typically bulk diffusion). Tritium diffusion
is slow at lower temperatures and leads to unacceptable tritium
inventory retained in the solid breeder
– Tmax, maximum temperature limit, to avoid sintering (thermal and
radiation-induced sintering) which could inhibit tritium release;
also to avoid mass transfer (e.g., LiOT vaporization)
• The limitations on allowable temperature window,
combined with the low thermal conductivity, place limits
on allowable power density and achievable TBR
Solid Breeder Concepts: Key Advantages and Disadvantages
Advantages
• Non-mobile breeder permits, in principle, selection of a coolant that avoids
problems related to safety, corrosion, MHD
Disadvantages
• Low thermal conductivity, k, of solid breeder ceramics
– Intrinsically low even at 100% of theoretical density (~ 1-3 W · m-1 · c-1 for ternary
ceramics)
– k is lower at the 20-40% porosity required for effective tritium release
– Further reduction in k under irradiation
• Low k, combined with the allowable operating “temperature window” for solid
breeders, results in:
– Limitations on power density, especially behind first wall and next to the neutron
multiplier (limits on wall load and surface heat flux)
– Limits on achievable tritium breeding ratio (beryllium must always be used; still
TBR is limited) because of increase in structure-to-breeder ratio
• A number of key issues that are yet to be resolved (all liquid and solid
breeder concepts have feasibility issues)
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
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
through design, leading to higher thermal efficiency.
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
Large MHD drag results in large
MHD pressure drop
Conducting walls
Insulated wall
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)
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
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 wall
– 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
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
WHAT IS FCI ?
•
FCI (Flow Channel Insert) is the key
element of the DCLL blanket concept
•
Both ITER and DEMO
•
Made of 5-10 mm SiCf/SiC composite
•
Pressure equalization openings (slot or
holes) to nearly eliminate primary stress.
Secondary (thermal) stress still exists
•
The main functions are:
- to reduce the MHD pressure drop
(electrical insulation);
- to reduce heat leakage into He (thermal
insulation);
- to separate hot PbLi (650C) from Fe
•
No serious feasibility issues have been
identified yet. However tailoring SiC
properties and fabrication of complex
shape FCIs is still an issue.
Pb-17Li exit temperature can be
significantly higher than the operating
temperature of the steel structure =>
High Efficiency
SiCf/SiC FCI REQUIREMENTS
•
SiC=1-100 S/m: 101-103 reduction of MHD pressure drop
•
kSiC=1-10 W/m-K: heat leakage is <10% of the total power (DEMO)
•
The optimal (SiC,kSiC)* is strongly dependent on the thermofluid MHD and
should be determined by design tradeoffs, taking into account:
- P (<1-2 MPa)
- heat leakage (<10-15% of the total power)
- temperature gradient (<150-200 K per 5 mm FCI)
- PbLi-Fe interface temperature (<470-500C)
•
Suggested (DEMO): kSiC~2 W/m-K; SiC~100 S/m
(S.Smolentsev, N.Morley, M.Abdou, MHD and Thermal Issues of the
SiCf/SiC FCI, FST, July 2006 )
* Only k and   (across the FCI) are important
FCI RELATED R&D
Material science
Thermofluid MHD
• Development of low-conductivity
grade 2-D woven SiCf/SiC with a
thin surface sealing layer to
avoid soaking of PbLi into pores
(e.g. using CVD)
• Improvement of crack resistance
• Reliable measurements of
SiCf/SiC properties at 300 to
800C, including effect of
irradiation
• Fabrication of complex shape
FCIs with pressure equalization
openings and overlap sections
• Effectiveness of FCI as
electrical and thermal insulator
• Pressure equalization (slot or
holes ?)
• Effect of FCI on flow balancing
in normal and abnormal
(cracked FCI) conditions
• Optimal location of the FCIs in
the module
THERMOFLUID MHD ANALYSIS
Reduction of MHD pressure drop by FCI
Radial temperature distribution in
the poloidal duct (DEMO)
kSiC=2 W/m-K
recommended
5 mm FCI (SiC~10) reduces the MHD
pressure drop by factor of 100
Heat transfer is strongly affected
by SiCvia flow modification
MHD PHENOMENA in DCLL 1
• Effectiveness of FCI
as electrical/thermal
insulator
• MHD pressure drop
and flow balancing
• Buoyancy effects
and 2-D MHD
turbulence, and
their effect on
thermal behavior of
the module
US DEMO DCLL blanket module
Intensive studies, including modeling and
experiment, are being conducted at UCLA
MHD PHENOMENA in DCLL 2
• The flow in a
poloidal duct is
strongly affected by
cross-sectional
currents
• Interaction of the
currents with a
toroidal field results
in a jet-type flow
Velocity profile and cross-sectional
currents High velocity jets
Flow in the gap
MHD PHENOMENA in DCLL 3
• Strong influence of
buoyancy effect and 2D MHD turbulence on
heat transfer
• Reduction of high
velocity jets due to
turbulent diffusion
• About 10-time increase
in effective thermal
conductivity
• Circulation motion
Reduction of high velocity jets
in a turbulent flow
Laminar
Circulation motion
Turbulent
Many liquid wall reactor concepts for high power
density were conceived & analyzed in APEX
 Many candidate liquids were studied: Li,
Sn-Li, Sn, Flibe and Flinabe
 Several liquid wall flow schemes were
conceived:
–
–
–
–
Thick liquid walls
Thin fast flowing protection layer (CLIFF)
Inertial or EM assisted wall adhesion
Integrated or stand-alone divertors Surface
 Concept performance was
analyzed from many perspectives
Fast Flow
Cassette
Inboard
Fast
Flow
Outboard
Fast Flow
Divertor
Cassette
Renewal
– Liquid wall flow MHD and heat transfer
– Breeding, shielding and activation potential
– Simplicity of system design, maintenance
 Interactions of liquid walls with plasma
operation were emphasized
Bottom Drain
Flow
– Plasma edge effects, impurities & recycling
– Liquid metal motion coupling to plasma
Thin liquid wall concept (blanket
modes
region behind LW not shown)
New simulation tools and experimental facilities
used to address flowing liquid metals in NSTX
divertor fields – now being applied to DCLL-TBM
 New phenomenon observed in both experiments and
numerical simulation for film flows in NSTX divertor: the
liquid film tends to ‘pinch in’ away from the wall under a
positive surface normal magnetic field gradient.
PbLi
FCI
‘Pinching in’
Gallium flow experiment at UCLA M-TOR facility
HIMAG simulation of the above experiment
Flow Velocity : 3 m/s
 Simulation with MHD research
code (at UCLA) shows tendency
for strong reversed flow jets near
slot or crack in flow channel
insert (MTOR experiments in
development)
Average surface normal field gradient: 0.6 T/m
HyPerComp Incompressible MHD solver for Arbitrary Geometry
The primary objective of HIMAG is to model the flow of liquid metals in
nuclear fusion reactor design. These flows are characterized by very high
magnetic field strength, complex geometry and fluid-solid coupling via
electric current, heat and mass transfer.
There is no other code (commercial or otherwise,) which is capable of
producing accurate and reliable solutions in such flows.
HIMAG: Technical Summary
 HIMAG is a parallel, unstructured mesh based MHD solver.
 High accuracy at high Hartmann numbers is maintained even on nonorthogonal meshes
 HIMAG can model single-phase as well as two-phase (free surface) flows
 Multiple conducting solid walls may be present in the computational
domain
 Graphical User Interfaces are provided for the full execution of HIMAG
 Heat transfer, natural convection, temperature dependent properties can
be modeled
Extensive validation and benchmarking has been performed for canonical
problems. Cases involving Ha > 1000 have never been
demonstrated on non-rectangular meshes prior to HIMAG
HIMAG: Single-phase flow studies
Cylinder flow with MHD,
Ha = 1000
Very high Hartmann
number (>10,000) computed
and verified
Natural convection with MHD
streamlines (above),
current lines (below), against
temperature contours
HIMAG: Two-phase flow studies
Plasma-liquid interaction
Liquid metal jet in a
magnetic field
Flows past obstacles, and
Complex geometries
Droplets and other 3-D phenomena
Flows with large deformation
HIMAG: Survey of applications
 HIMAG is being actively used to assist current research projects (ALPS,
ITER-TBM)
 HIMAG can readily be applied to metallurgical processes which use
magnetic fields e.g., continuous casting of Al and Steel
 HIMAG is being extended to model physical phenomena usable in
micro and nano flows where surface tension and capillarity
determine fluid motion and EM interactions are important
 HIMAG is being extended to model phase-change in the form of
boiling, and solidification. Various industrial processes require
this capability
 HyPerComp is developing a multi-physical simulation capability
with HIMAG being a key ingredient, in the modeling of
flows in nuclear engineering, and aerospace
(A “Virtual Test Blanket Module” using HIMAG is in the planning stage)
Summary
• The D-T Fusion process offers the promise of:
– Virtually unlimited energy source from cheap abundant fuels;
– No atmospheric pollution of greenhouse and acid rain gases;
– Low radioactive burden from waste for future generations.
• Tremendous Progress has been achieved over the past
decades in plasma physics and fusion technology.
• Fusion R&D involves many challenging areas of physics
and technologies and is carried out through extensive
international collaboration
• EU, JA, USA, RF, PRC, Korea, and India are about to
construct ITER to demonstrate the scientific and
technological feasibility of fusion energy (ITER will
produce 500MW of fusion power )