Transcript Fusion Nuclear Technology Grand Challenges with Exciting Opportunities for Young Researchers Mohamed Abdou Keynote Presentation at ISFNT-7, Tokyo, May 25, 2005

Fusion Nuclear Technology
Grand Challenges
with Exciting Opportunities
for Young Researchers
Mohamed Abdou
Keynote Presentation at ISFNT-7, Tokyo, May 25, 2005
Acknowledgements
• Heartfelt thanks to the ISFNT-7 organizers (particularly the
Chairman, Dr. Masahiro Seki) for their dedicated efforts in planning
a very successful conference.
Thanks also for the invitation to give this presentation.
• I solicited input on advances and challenges of FNT from
many experts and leaders of FNT around the world. I truly
appreciate their thoughtful and insightful responses.
Siegfried Malang
Alice Ying
Dai-Kai Sze
Takeo Muroga
Akio Sagara
Scott Willms
Brad Merrill
Phil Sharpe
Luciano Giancarli
Neil Morley
Shahram Sharafat
Akira Kohyama
Mohamed Sawan
Clement Wong
Dave Petti
Jeff Brooks
Masato Akiba
Lorenzo Boccaccini
Steve Zinkle
Sergey Smolentsev
Nasr Ghoniem
Paola Batistoni
Lee Cadwallader
Ahmad Hassanein
2
Fusion Nuclear Technology (FNT)
Fusion Power & Fuel Cycle Technology
FNT Components and Materials 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
3
Reflections on
Progress and Challenges of the Last Decade
• During the past 10 years, research on FNT made a lot of progress
in several areas, but also encountered some disappointments—
some issues proved more challenging than previously thought
• The World’s Fusion Effort Focused on ITER:
– Most extensive engineering effort to date
– ITER Engineering design clearly brought about the importance of FNT
– Despite the fact that power load, fluence and temperature in ITER are
modest, it took much time and effort to understand and conceive acceptable
solutions to FNT design issues (first wall, divertor, tritium, etc.)
Conclusions:
•
•
Successful ITER operation and an effective ITER Test Blanket Module (TBM)
still require important FNT R&D
FNT for devices beyond ITER is very challenging & requires focus & innovation
ITER is the Beginning, not the End
4
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 / Availability
7. Testing in Fusion Facilities
5
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
6
World Tritium Supply Would be Exhausted by 2025
if ITER Were to Run at 1000 MW at 10% Availability
Projected Ontario (OPG) Tritium
Inventory (kg)
30
25
CANDU Supply
20
w/o Fusion
15
1000 MW Fusion,
10% Avail, TBR 0.0
10
ITER-FEAT
(2006 start construction)
5
0
2000
•
•
2010
2020
Year
2030
2040
Availability of external tritium supply for continued fusion development beyond ITER first
phase is an issue
Large power D-T facilities must breed their own tritium
(this is why ITER’s extended phase was planned to include the installation of a tritium breeding blanket)
•
Blanket development and ITER-TBM are necessary in the near term to allow continued
development of D-T fusion
7
Tritium self-sufficiency condition:
Λa > Λ r
Λr = Required tritium breeding ratio
Λr is 1 + G, where G is the margin required to account
for tritium losses, radioactive decay, tritium inventory in
plant components, and supply inventory for start-up of
other plants.
Λr is dependent on many system physics and
technology parameters.
Λa = Achievable tritium breeding ratio
Λa is a function of technology, material and physics.
8
Λa = Achievable tritium breeding ratio
Λa is a function of technology, material and physics.
– FW thickness, amount of structure in the blanket, blanket concept.
30% reduction in Λa could result from using 20% structure in the blanket.
(ITER detailed engineering design is showing FW may have to be much
thicker than we want for T self sufficiency)
– Presence of stabilizing/conducting shell materials/coils for plasma
control and attaining advanced plasma physics modes
– Plasma heating/fueling/exhaust, PFC coating/materials/geometry
– Plasma configuration (tokamak, stellerator, etc.)
Integral neutronics experiments in Japan and the EU showed
that calculations consistently OVERESTIMATE experiments
by an average factor of ~ 1.14
Analysis* of current worldwide FW/Blanket concepts
shows that achievable TBR Λa ≤ 1.15
* See, for example, Sawan and Abdou (this conference)
9
Dynamic fuel cycle models were developed to calculate
time-dependent tritium flow rates and inventories
Such models are essential to predict the required TBR
(Dynamic Fuel Cycle Modelling: Abdou/Kuan et al. 1986, 1999)
Simplified Schematic of Fuel Cycle
To new
plants
Startup
Inventory
T storage and
management
Impurity separation
and
Isotope separation system
T waste
treatment
Fueling
system
DT
plasma
Exhaust Processing
(primary vacuum pumping)
T processing
for blanket
and PFC
depends on
design option
PFC
Blanket
10
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 [%]
11
Physics and Technology R&D needs to assess the
potential for achieving “Tritium Self-Sufficiency”
1. Establish the conditions governing the scientific
feasibility of the D-T cycle, i.e., determine the
“phase-space” of plasma, nuclear, material, and
technological conditions in which tritium selfsufficiency can be attained
– The D-T cycle is the basis of the current world plasma physics and
technology program. There is only a “window” of physics and
technology parameters in which the D-T cycle is feasible. We need
to determine this “window.” (If the D-T cycle is not feasible the
plasma physics and technology research would be very different.)
– Examples of questions to be answered:
–
–
–
–
Can we achieve tritium fractional burn-up of >5%?
Can we allow low plasma-edge recycling?
Are advanced physics modes acceptable?
Is the “temperature window” for tritium release from solid
breeders sufficient for adequate TBR?
– Is there a blanket/material system that can exist in this phasespace?
12
R&D for Tritium Self-Sufficiency (cont’d)
2.
Develop and test FW/Blankets/PFC that can operate in
the integrated fusion environment under reactorrelevant conditions
–
3.
R&D on FW/Blanket/PFC and Tritium Processing
Systems that emphasize:
–
–
–
4.
The ITER Test Blanket Module (TBM) is essential for
experimental verification of several principles necessary for
assessing tritium self-sufficiency
Minimizing Tritium inventory in components
“Much faster” tritium processing system, particularly processing
of the “plasma exhaust”
Improve reliability of tritium-producing (blanket) and tritium
processing systems
R&D on physics concepts that improve the tritium
fractional burn-up in the plasma to > 5%
13
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
14
Progress on High Power Density
Substantial progress has been made over the
past several years in exploring first wall /
blanket / divertor concepts with high power
density capability:
a) Liquid walls/liquid surfaces (mostly in the APEX
and ALPS studies)
b) Advanced solid first wall concepts (e.g., EVOLVE
and DCLL)
c) Advanced solid divertor concepts (especially in
EU)
15
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) 16
Some Key Points From Liquid Wall Studies
 Thin fast flowing layers protecting more conventional closed channel
blankets appear to be the most feasible and attractive concept
 high power density capability
 disruption survivability
 improved plasma performance
(Thick liquid walls for tokamaks appear very difficult to implement for a number
of reasons, especially MHD and flow control )
 Based on comprehensive plasma edge modeling studying impurity
vapor intrusion into core plasma, Liquid Sn and Sn-Li have the highest
surface temperature capability (> 630ºC)
 Flinabe salt with low melting point (~300ºC) is a promising alternative to LMs
 Liquid walls have strong possibilities to improve plasma performance




Close fitting conducting shell effects
Hydrogen gettering leading to low recycling
Impurity gettering
Helium trapping and pumping in nano-bubbles
(but “Rotating shell” effects on Resistive Wall Modes due to fast LM motion do
not appear to aid stabilization)
17
LW research has been a driver for
advances in plasma science and MHD
thermofluid modeling and experiments
Liquid wall research has:
 Stimulated considerable interest among plasma
physicists because of potentially large improvements in
plasma performance
 Plasma experiments conducted in PISCES, CDX-U, and DIII-D
 New experiments are planned for NSTX and the new LTX
 Modeling of plasma edge, plasma stabilization
 Motivated advances in modeling and experiments on
MHD effects on fluid flow, heat and mass transfer for
liquid metals and molten salts.
These advances have provided a much needed
capability for modeling and experiments of closed
channel blankets such as in ITER-TBM
18
Plasma Improvement in CDX-U with Liquid Lithium
• First fully-toroidal liquid lithium
limiter tested
• Mechanically stable in
magnetically confined plasmas
• Strong reduction in oxygen
• Strong reduction in edge fuel
recycling demonstrated by
dramatically higher fueling rates
needed to maintain plasma density
• Higher Te, higher current
3.5 1019
Total fueling (no. of part.)
Liquid lithium
tray limiter in
CDX-U
3 1019
Liquid lithium
plasma limiter
2.5 1019
2 1019
1.5 1019
Higher plasma
currents achieved
with liquid lithium
1 1019
(no lithium)
5 1018
0
30
40
50
60
70
80
Plasma Current (kA)
19
Lithium Tokamak eXperiment (LTX) and
NSTX lithium divertor experiments at PPPL
will test transport and plasma profile modification
with low-recycling lithium walls
LTX Conducting shell
Li Divertor flow module
for NSTX
LTX In-vessel conformal
chromium copper shell with
renewable layers of lithium
deposited by special
evaporators – Li area 4 m2
 NSTX, currently testing with Li pellets,
will move to coated surfaces to
investigate machine compatibility and
1st order lithium effects on plasma
operation
 Evaluating a flowing liquid divertor
module as a particle control solution 20
Recent PROGRESS in Modeling MHD Flows for Fusion
The starting point (several years ago):
• no commercial MHD codes;
• high Hartmann number codes limited to 2-D or 3-D inertialess flows;
• full 3-D research codes for geometrically simple flows limited to Ha~300-500.
• No real free surface simulation capability (few 2D fully developed or averaged
codes only)
Recent developments: (motivated by liquid wall research and ITER-TBM needs)
• HIMAG (HyPerComp Incompressible MHD Solver for Arbitrary Geometry) is a
new MHD software being developed by HyPerComp with support from UCLA. It
can be applied to both free surface and closed channel flows with a complex
geometry at high Ha (present limit is Ha~103–104).
•
2-D and 3-D research codes are being developed at UCLA in parallel with
HIMAG. The goal is to test and verify new models (e.g. MHD turbulence, MHD
natural convection, tritium transport, etc.) as well as new numerical approaches,
and then incorporate them in the HIMAG.
•
Modified commercial codes (FLUENT, CFX, FLOW3D), which are capable to
model some MHD flows
21
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
22
Innovative Solid First Wall Concepts
EVOLVE (APEX)
- Novel concept based on use of high
temperature refractory alloy (e.g. tungsten)
with innovative heat transfer/transport
scheme for vaporization of lithium
- Low pressure, low stresses
- Low velocity, MHD insulator not required
- High power density / temperature /
efficiency
- Key issues relate to
tungsten
• Attempts to extend the capabilities and attractiveness of solid walls
have required very advanced structural materials
• EVOLVE requires W alloy for high power density, high temperature
But the Material Community is not enthusiastic (risky, costly, very long-term)
23
High Power Density Divertor Plates
• Fusion reactors require divertor targets with
heat load capability > 10 MW/m2
• The ITER divertor concept (water-cooled copper heat sink) is
not attractive for reactors because:
– Copper’s lifetime is limited
– Water safety issues
(high chemical reactivity with liquid metal breeder and beryllium)
– Low-temperature water “wastes” divertor heat
• Earlier helium-cooled designs were limited to ≤ 5 MW/m2
• New development program in EU, starting around 2002, has
developed a promising helium-cooled divertor plate concept
with surface heat flux capability > 10 MW/m2
– Uses tungsten structure at > 800ºC (above embrittlement temperature)
– Helium at ~600–700ºC, suitable for high thermal efficiency
24
PL FUSION
Forschungszentrum Karlsruhe
in der Helmholtz-Gemeinschaft FZK - EURATOM ASSOCIATION
Helium-Cooled Divertor Plate with W can handle > 10 MW/m2
He
Armor (W)
Cap
(W Alloy)
Transition
Piece

STEEL
Cartridge
(Steel)
• High pressure helium (10 MPa)
• Novel ideas for increasing heat
transfer
He
600°C
He
700°C
FZK
02/24/2005
Features:
• Subdivide the divertor plate into
small modules to lower thermal
stresses: Plasma-facing surface
with ~20–30 mm
Results:
• Feasible, attractive design
• Heat flux > 10 MW/m2
• Helium at 600–700ºC → high
thermal efficiency
• Tungsten at > 800ºC (above
DBTT)
T. Ihli
25
Pathways to Higher Temperature
Operation in Practical FW/Blanket
• Operating at high temperature is needed for
higher thermal efficiency
• Advances were pursued on two different paths:
– Develop new high-temperature structural
materials
– Explore novel design concepts that can achieve
higher coolant temperature (> 650ºC) with present
generation structural materials which are
temperature-limited (< 550ºC)
• Results show both encouraging advances and
disappointing failures
26
Material Technology for fabricating high quality products
was significantly enhanced for V-alloy and SiC/SiC
(Large heats produced in the US and Japan, and then followed by Russia,
showed similar properties enhanced reliability of preparing V-alloys.)
High Purity V-4Cr-4Ti (NIFS-HEAT)
195 x 195 x 2 mm
F 31 mm
High Density SiC/SiC (NITE-Process)
[1] T. Muroga, M. Gasparotto and S. J. Zinkle, Fus. Eng. Des. 61-62 (2002) 13-25.
[2] A. Kohyama, K. Abe, A. Kimura, T. Muroga and S. Jitsukawa, Proc. TOFE-16.
27
Primary Damage Formation is Similar
for Fission and Fusion Neutrons
Stoller, Zinkle, ORNL
28
Development of Nano-Composited and
ODS Ferritic-Martensitic Steels
• These newly developed
alloys contain a high density
(~1024/m3) of nanoscale
particles formed from Y2O3
during processing.
New 12YWT Nano-composited
Ferritic Steel with Superior Strength
Compared to ODS Steels
• The thermal creep time to
failure is increased by
several orders of magnitude
around 800°C
PROMISE:
The new alloys offer the
potential for increasing the
operating temperature of iron
alloys by ~200°C
Zinkle, ORNL
29
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
30
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)
31
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
32
Effectiveness of SiCf/SiC FCI as
Electric/Thermal Insulator
X (poloidal)
FCI
Pb-17Li bulk flow
He flows
Fe wall
Pressure equalization opening makes the
pressure on both sides of the FCI equal,
resulting in almost no primary stresses. The
opening is either a slot (PES) or a row of
holes (PEH).
500
PES
(dP/dx)0 / (dP/dx)
Pb-17Li gap flow
B-field
PEH
PES
400
300
200
100
0
1
Minimizing heat leakage into He
by FCI allows for high Pb-17Li
bulk temperature of 650-700C,
while the He cooled Fe walls are
below 500 C.
10
100
1000
Electrical conductivity, S/m
Analysis shows significant MHD pressure drop
reduction with FCI as compared to the case
without insulation:
a factor of 10 at  = 500 S/m;
33
a factor of 200-400 at  = 5 S/m.
Dual Coolant PbLi (DCLL) Concept was Proposed in
Several European and US DEMO and Power Plant Studies
ARIES-ST Unit Cell
for DCLL Blanket
EU DCLL DEMO Blanket
ITER-TBM:
• The US selected the DCLL concept as the primary liquid breeder
concept for ITER-TBM
• Many of the ITER Parties are collaborating with the US on the DCLL.
34
Flow Channel Insert properties and
failures critically affect thermofluid MHD
performance potential

Electrical and thermal conductivity of the SiC/SiC perpendicular to
the wall should be as low as possible to reduce pressure drop and
to avoid velocity profiles with side-layer jets and excess heat
transfer to the He-cooled structure.

The inserts have to be compatible with Pb-17Li at temperatures up
to ~800 °C

Liquid metal must not “soak” into pores of the composite in order
to avoid increased electrical conductivity. In general, closed
porosity and/or dense SiC layers are required on all surfaces of the
inserts.

There are minimum primary stresses in the inserts. However,
secondary stresses caused by temperature gradients must not
endanger the integrity of irradiated FCIs.

The insert shapes needed for various flow elements must be
fabricable
35
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 or military
reactors (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.
36
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!
37
Tritium Permeation Barrier Development in EU
• Tritium permeation
barrier
development is a
key to tritium
leakage and
inventory control
• Development and
tests of tritium
permeation barrier
coatings (up to
2003, in the EU)
have not yet been
conclusive
Comparisons of permeability of HDA (Hot Dip
Aluminization) coated tubes in H2 gas and Pb-17Li
38
Key R&D Items for Tritium Control
Test Blanket Modules (TBMs) in ITER (and DT operation in
ITER) will give us the first quantitative real tests of the tritium
control and management issue.
Key R&D required toward successful demonstration:
• Sophisticated modeling tools capable of predicting the T-flows in
different blanket system and reactor components
– accounting for complexities from geometric factors, temperature dependent
properties, convection effects
• Continue to develop high performance tritium diffusion barrier
and clarify the still existing technological questions:
– understanding the sensitivity of the PRF to the quality of coating
– crack tolerance and irradiation experiments on coatings
– compatibility studies of coatings in flowing conditions at elevated
temperatures
• Continue to develop efficient tritium recovery system for both the
primary and the secondary coolants
– Efficiency to 99.99%
• Develop instrument capable of detecting tritium on-line down to a
very low concentration
39
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
40
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
41
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.
FNT Testing in Fusion Facilities
• ITER operation and conducting the Test Blanket Module
(TBM) Program in ITER will provide the first real
experimental results on the performance and issues of FNT
components and materials in the integrated fusion
environment.
• But are ITER tests sufficient to proceed to DEMO?
- Technical studies say other fusion test facilities (e.g. VNS/CTF)
are necessary.
- But official plans of some Parties still show ITER as the only fusion
facility from now to DEMO.
• Probably the biggest issue the International Community
needs to seriously address is how many major fusion
FNT experimental devices are needed to develop
practical fusion energy by the middle of the century.
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Type of Integrated Facility Needed for FNT Development
(blanket/FW, PFC, materials, tritium, safety)
in addition to ITER
• Need fusion environment, i.e., plasma-based facility
• Testing requirements are: (see IEA study)
– NWL > 1 MW/m2, steady state, test area ~ 10 m2, test volume ~ 5 m3
– Fluence requirements > 6 MW•y/m2
(engineering feasibility: 1–3 MW•y/m2; reliability growth > 4 MW•y/m2)
• What is needed is:
small power < 100 MW fusion power
long fluence > 6 MW•y/m2
• There is no external supply of tritium to run a large-power
device such as ITER for such fluences
• A device with small fusion power (~100MW) and
moderate wall load (~1 MW/m2) is a driven-plasma device
(Q~2). This is often called VNS or CTF.
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Summary Remarks
• The existence of the ITER detailed engineering design, ready
for construction, based on a large R&D effort, is the greatest
advance for FNT
• Conducting an effective Test Blanket Module (TBM) program is
one of the principal objectives of ITER
• ITER is the Beginning, not the End
• There are many remaining challenging FNT issues that need to
be resolved for successful fusion development
• Availability of external tritium supply for continued fusion
development beyond ITER’s first phase is an issue
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Summary Remarks (cont’d)
• “Tritium self-sufficiency” is a complex issue that depends on many
system physics and technology parameters / conditions.
We need to establish the physics and technology conditions governing
the scientific feasibility of the D-T cycle.
• Remarkable advances have been made in developing concepts with
high power density capability
– Thin liquid walls / liquid surfaces can handle high power density and have
positive impact on plasma performance
– He-cooled solid divertor plate with W can handle > 10 MW/m2
• The most challenging blanket and material problems are at the
INTERFACES: MHD insulators, thermal insulators, corrosion/
interface temperature limits, tritium permeation
Our progress in solving the interface problems has not been satisfactory
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Summary Remarks (cont’d)
• Advances in developing high-temperature structural
materials were made; but such advances cannot be utilized
unless we solve the interface problems.
• Innovative design concepts, such as dual-coolant PbLi, are
promising in providing high coolant temperature (> 650ºC)
while using existing ferritic steel, which has a temperature limit
of < 550ºC.
• Tritium control and management is a very difficult issue that
needs much more attention.
• Current plasma confinement configurations with long MTTR
and large wall area lead to reliability requirements on FNT
components that are more demanding than for other non-fusion
technologies.
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FNT research has made considerable progress, but
also has encountered some disappointments.
Resolving the remaining challenging FNT issues will
require “ingenuity”, particularly of young researchers.
FNT needs to attract and train bright young
scientists and engineers.
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