Transcript Overview of Principles, Concepts, and Key Issues of Fusion Nuclear Technology Mohamed Abdou Professor of Engineering and Director of Fusion Science and Technology Center University of.

Overview of Principles, Concepts,
and Key Issues of
Fusion Nuclear Technology
Mohamed Abdou
Professor of Engineering and
Director of Fusion Science and Technology Center
University of California Los Angeles
Seminar Presented to KAERI and KBSI, Korea, April 2004
Overview of Fusion Nuclear Technology (FNT)
Seminar Outline
Principles, Concepts, and Key Issues (this presentation/handout)
• FNT components and functions
• Tritium Breeding
• Solid Breeders
- Design Concepts and Materials
- Tritium Release, Extraction
- Issues and R&D
• Structural Materials
• Liquid Breeders
- Concepts: Materials, Configurations
- Li/V
- LiPb Dual Coolant
- Issues
• Tritium Supply and Need for Blanket Testing in ITER
• Possible Areas for US-Korea Collaboration
Other Presentations / Handouts
- Fabrication Technology
- ITER Test Blanket Module (TBM)
- ITER Testing : Engineering Scaling
- Molten Salts
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
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
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
ARIES-AT
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
Notes on FNT:
• The Vacuum Vessel is outside the
Blanket (/Shield). It is in a lowradiation field.
• Vacuum Vessel Development for
DEMO should be in good shape
from ITER experience.
• The Key Issues are for
Blanket / PFC.
• Note that the first wall is an
integral part of the blanket (ideas
for a separate first wall were
discarded in the 1980’s). The
term “Blanket” now implicitly
includes first wall.
• Since the Blanket is inside of the
vacuum vessel, many failures
(e.g. coolant leak from module)
require immediate shutdown and
repair/replacement.
Adaptation from ARIES-AT Design
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)
Poloidal Variation of Neutron Wall Load
– Neutron wall load has profile along the poloidal direction (due to combination of
toroidal and poloidal geometries)
–Peak to average is typically about 1.4
Inboard
(equatorial plane, outboard, in 0º)
outboard
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
1
(n;n’a) t
0.01
1
10
100
1000
10
4
10
Neutron Energy (eV)
5
10
6
Li  n  t  a  4.78MeV
7
Li  n  t  a  n  2.47MeV
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
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
Achievable TBR is Very Sensitive to FW Thickness
1.25
ARIES-ST
7% FS, 7% He,
12% SiC inserts, 74% LiPb
Overall TBR
1.20
1.15
Required TBR
1.10
1.05
0
0.5
1
1.5
Effective Thick. of FS in OB FW (cm)
2
ITER FW Panel Cross Section
TBR drops by up to 15% with 2 cm thick FS FW
[L. El-Guebaly, Fusion Engr & Design, 2003]
TBR is Very Sensitive to Structure Content in Blanket
2.0
1.5
Li 2O Breeder
1.8
1.4
V
TBR
TBR
1.6
V
FS
1.3
0
5
10
FS
SiC
1.4
1.2
SiC
1.2
Li 17Pb 83 Breeder
(90% Li-6)
15
20
1.0
0
5
10
15
20
Structure Content (%)
Structure Content (%)
2.0
 Impact of structure content on TBR
depends on breeder and structural
material used
Li Breeder
1.8
TBR
1.6
V
 V has the least impact on breeding
FS
1.4
SiC
 Up to 30% reduction in TBR could result
from using 20% structure in the blanket
1.2
1.0
0
5
10
15
Structure Content (%)
20
Note: Net TBR is substantially lower
(~30-40%) than local TBR
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)
JA Water-Cooled Solid Breeder Blanket
Neutron Multiplier
Be, Be12Ti (<2mm)
Tritium Breeder
Li2TiO3, Li2O (<2mm)
Optional W coating for
FW protection
Coolant water
(25MPa, 280/510oC)
MW/m2
Surface Heat Flux:1
Neutron Wall Load: 5 MW/m2(1.5×1015n/cm2s)
First Wall
(RAFS, F82H)
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
Some mathematical formulas
Diffusion model:
Generation rate
  C (r , t ) 2 C (r , t ) 
C (r , t )
  G(r , t )
 D(T )

2
t
r r 
 r
2
Activation
D(T )  D0 exp(Ed / RT )
energy
C (r ,0)  0
C ( a, t )  0
 C (r , t ) 

 0
 r  r 0
3
n
G 2
2
Ga
(

1
)
 nr 
2
2 2
2
C
(a  r ) 
sin
x
exp

Dn

t
/
a


3 
6D
D r
n3
 a 

First -order tritium release rate estimated:
Surface concentration (atoms/m2)
R(t )  dCs / dt  K des (t )C s (t )  K 0 C s (t ) exp( E des / RT (t ))
t
rate constant
 Desorption

C s (t )  C s 0 exp K 0  exp( E des / RT (t ' ))dt'
 0

Desorption energy

MISTRAL (Model for Investigative Studies of Tritium
Release in Lithium Ceramics)- a code developed at UCLA
To understand and predict tritium
release characteristics
Gas phase
Solid phase
Phenomenological cartoon
Transport mechanisms included:
grain diffusion
grain boundary diffusions
adsorption from the bulk and
from the pores to the surface
desorption to the pores
diffusion through the pores
Features
• includes details of the ceramic
microstructure
• includes coverage dependence
of the activation energy of
surface processes (adsorption/
desorption)
“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
Effect of helium purge flow rate on pressure drop and tritium permeation
P  175
(1  a ) 2
 f NRTL
a3
(d p )2 Ab ( P0  P / 2)
a =Porosity,
 = pebble sphericity =1 for spherical
pebble
N = moles/s
R = ideal gas constant
T = temperature
single size bed
binary bed
Porosity, a
f = helium gas viscosity
Ab= gas flow cross-sectional area
P0= inlet pressure
L = flow path
dp = particle diameter
Which solid breeder ceramic is better?
Parameters:
Lithium density
Tritium residence time
Thermal-physical properties
Mechanical properties
Temperature window
Transmutation nuclides (activation
products)
Reactivity
Fabrication
Irradiation effects (e.g, swelling)
Notes:
• Li2O is highly hygroscopic:
2Li2O + H2O → 2LiOH (ΔH =
128.9 kJ/mole); LiOH is highly
corrosive
• Li2O has been observed to
swell under irradiation
• Li2O is the only ceramic that
may achieve the desired TBR
without a neutron multiplier
(but not assured)
Properties are for
100% TD
Li2O Li4SiO4 Li2TiO3
Li2ZrO3
Lithium Density (g/cm3)
0.94
0.51
0.43
0.38
Diameter (mm)
~1.0
0.2~0.7
0.7~0.85
0.9~1.5
Thermal Expansion
@ 500 ° C (L/L0%)
1.25
1.15
0.8
0.5
Thermal Conductivity
4.7
2.4
1.8
@ 500 ° C (W/m/ ° C) Higher design margin
0.75
Min.-Max. Temp. for
Tritium Release (°C)
397795
325-925 Up to 900 400-1400
Relatively narrow T window
Swelling @500 ° C
7.0
1.7
-
< 0.7
Reactivity w/H20
High
Little
Less
Less
Grain Size (μm)
50
5-15
1-4
0.5-2
80-85
~98
87~89
93~96
Crush Load (N)
-
~ 10
24-33
68-79
Residence time @400 °C (h)
10
(V/V0%)
Pore for
Density (%TD) tritium
release
2
2
1
Tritium Extraction for Solid Breeder Blankets
1% H2
Absorb impurities from tritium
stream at ambient temperature
Remove the remaining
impurities and the
hydrogen isotopes at a
cryogenic temperature
molecular sieve bed
Tritium form:
HT and HTO
The “bleed” stream is sent to a shift catalyst bed where
reactions such as steam reforming and water gas shift
can be used to move hydrogen isotopes from impurities
such as CQ4 and Q2O to the form of Q2
Legend
AMSB
CMSB
CR
ISS
TWT
Ambient Molecular Sieve Bed
Cryogenic Molecular Sieve Bed
Catalytic Reactor
Isotope Separation System
Tritium Waste Treatment
When the CMSB is
saturated with Q2
(hydrogen isotopes) it is
taken off line for
regeneration and its
companion bed can be put
into service. A CMSB is
regenerated by warming.
The Q2 desorbs and is sent
to a Pd/Ag permeator.
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)
Configurations and Interactions among breeder/Be/coolant/structure are
very important and often represent the most critical feasibility issues.
• Configuration (e.g. wall parallel or
“head on” breeder/Be arrangements)
affects TBR and performance
• Tritium breeding and release
- Max. allowable temp. (radiationinduced sintering in solid breeder
inhibits tritium release; mass
transfer, e.g. LiOT formation)
- Min. allowable Temp. (tritium
inventory, tritium diffusion
- Temp. window (Tmax-Tmin) limits
and ke for breeder determine
breeder/structure ratio and TBR
• Thermomechanics interactions of
breeder/Be/coolant/structure involve
many feasibility issues (cracking of
breeder, formation of gaps leading to
big reduction in interface conductance
and excessive temperatures)
Thermal creep trains of Li2TiO3 pebble bed at
different stress levels and temperatures
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
Major R&D Tasks for Solid Breeder Blanket
• Solid breeder material development, characterization, and fabrication
• Multiplier material development, characterization, and fabrication
• Tritium inventory in beryllium; swelling in beryllium irradiated at
temperature, including effects of form and porosity
• Breeder and Multiplier Pebble Bed Characterization
• Pebble bed thermo-physical and mechanical properties,
thermomechanic interactions
• Blanket Thermal Behavior
•
•
•
•
•
Neutronics and tritium breeding
Tritium Permeation and Processing
Nuclear Design and Analysis (Modeling Development)
Advanced In-Situ Tritium Recovery (Fission Tests)
Fusion Test Modules Design Fabrication and Testing
• Material and Structural Response
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
Material
Coolant/Tritium Breeding Material
Li/Li
Ferritic steel
V alloy
SiC/SiC
He/PbLi
H2O/PbLi
He/Li ceramic H2O/Li ceramic FLiBe/FLiBe
Comparison of fission and fusion structural
materials requirements
Fission
(Gen. I)
Fission
(Gen. IV)
Fusion
(Demo)
Structural alloy maximum
temperature
<300˚C
600-850˚C
(~1000˚C for
GFRs)
550-700˚C
(1000˚C for SiC)
Max dose for core internal
structures
~1 dpa
~30-100 dpa
~150 dpa
Max transmutation helium
concentration
~0.1 appm
~3-10 appm
~1500 appm
(~10000 appm for
SiC)
• Fusion has obtained enormous benefits from pioneering radiation effects research
performed for fission reactors
– Although the fusion materials environment is very hostile, there is confidence that
satisfactory radiation-resistant reduced activation materials can be developed if a
suitable fusion irradiation test facility is available
Fission (PWR)
Fusion structure
Coal
Tritium in fusion
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.
Physical Properties of Molten Natural Li (temperature in degrees Kelvin)
Valid for T = 455-1500 K
Melting Temperature: 454 K (181ºC)
Density [1]
r (kg/m3) = 278.5 - 0.04657 · T + 274.6 (1-T/3500)0.467
Specific heat [1; see also 2]
CP (J/kg-K) = 4754 - 0.925 · T + 2.91 x 10-4 · T2
Thermal conductivity [1]
Kth (W/m-K) = 22.28 + 0.0500 · T - 1.243 x 10-5 · T2
Electrical resistivity [1]
re (nWm) = -64.9 + 1.064 · T - 1.035 x 10-3 T2 + 5.33 x 10-7 T3 - 9.23 x 10-12 T4
Surface tension [1]
g (N/m) = 0.398 - 0.147 x 10-3 · T
Dynamic viscosity [1] note: h = ru where u = kinematic viscosity (m2/s)
ln h (Pa - s) = -4.164 - 0.6374 ln T + 292.1/T
Vapor pressure [1]
ln P (Pa) = 26.89 - 18880/T - 0.4942 ln T
References:
[1] R.W. Ohse (Ed.) Handbook of Thermodynamic and Transport Properties of Alkali Metals, Intern. Union
of Pure and Applied Chemistry Chemical Data Series No. 30. Oxford: Blackwell Scientific Publ., 1985,
pp. 987.
[2] C.B. Alcock, M.W. Chase, V.P. Itkin, J. Phys. Chem. Ref. Data 23 (1994) 385.
Physical Properties of Pb-17Li
Melting Temperature: TM = 507 K (234ºC)
Density [1]
r (kg/m3) = 10.45 x 103 (1 - 161 x 10-6 T)
508-625 K
Specific heat [1]
CP [J/kg-K] = 195 - 9.116 x 10-3 T
508-800 K
Thermal Conductivity [1]
Kth (W/m-K) = 1.95 + 0.0195 T
508-625 K
Electrical resistivity [1]
re (nW-m) = 10.23 + 0.00426 T
508-933 K
Surface tension [2,3]
g(N/m) =0.52 - 0.11 x 10-3 T
520-1000 K
Dynamic viscosity [1]
h (Pa - s) = 0.187 x 10-3 exp [1400./T]
521-900 K
Vapor pressure [2-4]
P (Pa) = 1.5 x 1010 exp (-22900/T)
550-1000 K
References:
[1] B. Schulz, Fusion Eng. Design 14 (1991) 199.
[2] H.E.J. Schins, Liquid Metals for Heat Pipes, Properties, Plots and Data Sheets, JRC-Ispra (1967)
[3] R.E. Buxbaum, J. Less-Common Metals 97 (1984) 27.
[4] H. Feuerstein et al., Fusion Eng. Design 17 (1991) 203.
Physical Properties of Molten Flibe (LiF)n · (BeF2)
Melting temperature [1]
TM(K) = 636 K (363ºC)
TM(K) = 732 K (459ºC)
n=0.88
n=2
(TM=653 K for n=1)
Density [2]
r (kg/m3) =2349 – 0.424 · T
r (kg/m3) =2413 – 0.488 · T
n=1
n=2
930-1130 K
800-1080 K
Specific heat [3]
CP (J/kg-K) ≈ 2380
n=2
600-1200 K ?
Thermal conductivity [3]
Kth (W/m-K) = 1.0
n=2
600-1200 K ?
Electrical resistivity [2]
re (W-m) = 0.960 x 10-4 exp (3982/T)
re (W-m) = 3.030 x 10-4 exp (2364/T)
n=1
n-2
680-790 K
750-920 K
Surface tension [2,4]
g (N/m) = 0.2978 - 0.12 x 10-3 · T
g (N/m) = 0.2958 - 0.12 x 10-3 · T
n=1
n=2
830-1070 K
770-1070 K
Dynamic viscosity [2]
h(Pa - s) = 6.27 x 10-6 exp (7780/T)
h(Pa - s) = 5.94 x 10-5 exp (4605/T)
n=1
n=2
680-840 K
740-860 K
Vapor pressure [3]
P (Pa) = 1.5 x 1011 exp (-24200/T)
n=2
770-970 K
References:
[1]
K.A. Romberger, J. Braunstein, R.E. Thoma, J. Phys. Chem. 76 (1972) 1154.
[2]
G.J. Janz, Thermodynamic and Transport Properties for Molten Salts: Correlation equations for critically evaluated density, surface tension, electrical
conductance, and viscosity data, J. Phys. Chem. Ref. Data 17, Supplement 2 (1988) 1.
[3]
S. Cantor et al., Physical Properties of Molten-Salt Reactor Fuel, Coolant and Flush-Salts, ORNL-TM-2316 (August 1968).
[4]
K. Yajima, H. Moriyama, J. Oishi, Y. Tominaga, J. Phys. Chem. 86 (1982) 4193.
Liquid Breeders
Summary of some physical property data
• Some key physical property data for Flinabe are not yet available
– (melting temperature measurements for promising compositions are in progress.
Measurement at Sandia in early 2004 shows ~ 300ºC)
• Physical property data for Flibe are available from the MSR over a
limited temperature range
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
r
r
– 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)
LM-MHD pressure drop window for
inboard channels is closed!
( NWL ) L2 B 2 w
S
rc p T
Lithium Inboard Base Case
(Sze, 1992)
- NWL = 5 MW/m2
- Blanket Thickness = 0.2 m
- Blanket Length = 6.0 m
ITER FW
ARIES-RS FW - Coolant Bulk T rise = 200 K
Pipe Stress (MPa)
300
250
Base Case
200
Blanket Thickness = .25 m
150
Flow Length = 4 m
100
Bulk T rise = 300 K
50
NWL = 2.5 MW/m2
0
1
2
U ~ .2-.3 m/s
3
4
5
6
7
8
9 10 11 12 13 14 15
Magnetic Field (T)
So a strategy is needed to
reduce MHD pressure drop for liquid metals
PMHD  KL lUB 2
“K” factor represents a measure of relative
conductance of induced current closure paths
• Lower K
Main options considered: Break
– Insulator coatings/Laminated walls electrical coupling to load bearing

– Flow channel inserts
walls so pipe walls can be made
thick for more strength without
– Elongated channels with anchor
also increasing pressure drop!
links or other design solutions
• Lower Velocity: U
– Heat transfer enhancement or dual/separate coolant to lower
velocity required for first wall/breeder zone cooling
– High temperature difference operation to lower mass flow
• Lower Magnetic field and flow length: B,L
– Outboard blanket only, with poloidal segmentation
• Lower electrical conductivity:  (molten salt)
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 5m/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
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
He-Cooled PbLi Flow Scheme
• PbLi is fed at the top and
collected at the back
• Meandering PbLi flows in vertical
columns delimited by vertical SPs
• Alternative flow holes at
front/back of horizontal SPs
[18-54] mm/s
PbLi inlet
[0.5-1.5] mm/s
pol
rad
PbLi outlet
Key features of Dual Coolant Lead-Lithium Concept
(One of the concepts considered by the U.S. for ITER TBM)
• Cool the ferritic steel FW and structure with separate
coolant – He (also used for FW/blanket preheating and possible tritium control)
– The idea is to keep the structure temperature below 550ºC, the allowable
temperature for ferritic steel, and the interface temperature below 480ºC
• Breeding zone is self-cooled PbLi
– PbLi can be moving at slow velocity, since the heat generation rate in the breeding
zone is lower than the surface heat flux at the FW
– PbLi can be operated at temperatures higher than the structure, for higher
thermodynamic efficiency.
– Some type of thermal/MHD insulator, e.g., FCIs, is required, but the requirements
are more relaxed than for “all self-cooled” concepts
• Use flow channel inserts (FCIs), wherever possible to:
– Provide electrical insulation to reduce MHD pressure drop
– Provide thermal insulation to decouple PbLi bulk flow temperature from wall
temperature
– Provide permeation barrier to reduce T permeation into the He system
– Provide additional corrosion resistance since only stagnant PbLi is in contact with
the ferritic steel structural walls
Dual Coolant Concept Designs from EU and USA
Cross section of the breeder region unit cell
(ARIES)
Flow Channel Insert Properties and Failures are
Dominant Issues for PbLi Dual Coolant Blankets
•
Electrical and thermal conductivity of the SiC/SiC should be as low as
possible 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
700-800 °C
•
Liquid metal must not “soak” into pores of the composite in order to
avoid increased electrical conductivity and high tritium retention. In
general “sealing layers” are required on all surfaces of the inserts.
– even if the change in conductivity is modest from pressure drop
point of view it could also affect flow balance
•
There are minimum primary stresses in the inserts. However,
secondary stresses caused by temperature gradients must not
endanger the integrity under high neutron fluence.
•
The insert must be applicable and affordable
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 dual-coolant concepts)
Molten Salt Blanket Concepts
(One of the concepts considered by U.S. for ITER TBM)
• Lithium-containing molten salts are used as the coolant fot 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
Concepts considered by US for ITER TBM:
– Dual coolant (He-cooled ferritic structures, self-cooled molten salt)
– Self-cooled (only with low-melting-point molten salt)
Dual Coolant Molten Salt Blanket Concepts
• He-cooled First Wall and structure
• Self-cooled breeding region with flibe or flinabe
• No flow-channel insert needed (because of lower conductivity)
Example: Dual-Cooled FLiBe + Be Blanket Concept
Helium
Flows
Poloidal cross-section
Helium
Flows
Self-cooled – FLiNaBe Design Concept
Radial Build and Flow Schematic
FLINaBe Out
2/3
FLINaBe Out
1/3
FLINaBe In
Tritium Extraction from Liquid Lithium
The liquid lithium exits the torus
and has protium added to it.
This hydrogen-swamped stream is
introduced into a cooler to reduce
the temperature to 200 C. This will
cause a portion of LiH and LiT to
precipitate from the liquid lithium.
This two-phase mixture is then
sent to a cold trap where the
LiQ (Q represents H, D and T) is
collected. This process reduces
the tritium concentration in the
stream to 1 ppm.
Lastly, the liquid lithium and
some of the LiQ is sent to a
heater before being
reintroduced into the torus.
Periodically the cold trap will require
regeneration. This can be accomplished by
heating the cold trap to 600 C. At this
temperature the hydrogen isotopes will exert
a 10 torr partial pressure which can be
pumped away and recovered.
Tritium Removal and Recovery from LiPb
Technique includes the following
steps:
• tritium permeation into the
NaK-filled gap of the doublewalled heat exchanger
• tritium removal from the NaK
by precipitation as potassium
tritide in a cold trap
• tritium recovery by thermal
decomposition of the tritide
and pumping off the tritium
gas
Two cold traps are operated in parallel: one for tritium removal by circulating the
tritium dissolved in the NaK to the cold trap; the other for tritium recovery. For
this purpose the cold trap is decoupled from the circulation loop, drained from NaK,
heated up to temperatures of about 380oC and the released tritium gas is pumped
off and stored in a getter bed.
Tritium Consumption and Production
Fusion Consumption
55.8 kg per 1000MW fusion power per year
Production & Cost
• CANDU Reactors: 27 kg over 40 years, $30M/kg (current)
• Fission reactors: few kg per year, $200M/kg!! (projected cost
after Canadian tritium is gone) It takes tens of fission reactors
to supply one fusion reactor.
Conclusions
• ITER’s extended phase requires tritium breeding.
• Large power DT facilities must breed their own tritium.
World Tritium Supply Would be Exhausted by 2025
if ITER Were to Run at 1000MW at 10% Availability
(OR at 500 MW at 20% 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
(2004 start)
5
0
1995
2000
2005
2010
2015
2020
Year
2025
2030
2035
2040
2045
Tritium supply and self-sufficiency are as critical to
fusion energy as demonstrating a burning plasma.
They are “Go-No Go” Issues for Fusion:
– There is no practical external source of tritium for
fusion energy development beyond a few months of DT
plasma operation in an ITER-like device.
– There is NOT a single experiment yet in the fusion
environment to show that the DT fusion fuel cycle is
viable.
• Early development of tritium breeding blanket is
critical to fusion now
• Testing breeding blanket modules in ITER is
REQUIRED
Testing in a Fusion Facility is the fastest approach to Blanket and
Fusion Development to Demo
A fusion test facility allows SIMULTANEOUS testing of integrated
(synergistic) effects, multiple effects, and single effects
- Allows understanding through single and multiple effects tests under same conditions
- Provides “direct” answer for synergistic effects
Specimen (thousands)
100 cm
50 cm
9 cm
2.5 cm
10.8 cm
Capsule test (100’s)
* Figures are not to scale. Note Dimensions
Submodule
(>100)
Test Module
(>30)
• Also Test
Sectors (several)
ITER Provides the First Integrated Experimental
Conditions for Fusion Technology Testing
• Simulation of all Environmental Conditions
Neutrons
Plasma Particles
Electromagnetics
Tritium
Vacuum
Synergistic Effects
• Correct Neutron Spectrum (heating profile)
• Large Volume of Test Vehicle
• Large Total Volume, Surface Area of Test Matrix
But ITER Operating Parameters pose a serious
challenge
to obtaining meaningful blanket testing
results. Careful design of ITER test blanket module
(TBM) must be based on detailed engineering scaling.
Testing tritium breeding blankets has always
been a principal objective of ITER
• “The ITER should serve as a test facility for neutronics,
blanket modules, tritium production and advanced plasma
technologies. The important objectives will be the
extraction of high-grade heat from reactor relevant blanket
modules appropriate for generation of electricity.”
—The ITER Quadripartite Initiative Committee (QIC), IEA Vienna 18–19
October 1987
• “ITER should test design concepts of tritium breeding
blankets relevant to a reactor. The tests foreseen in
modules include the demonstration of a breeding
capability that would lead to tritium self sufficiency in a
reactor, the extraction of high-grade heat and electricity
generation.”
—SWG1, reaffirmed by ITER Council, IC-7 Records (14–15 December
1994), and stated again in forming the Test Blanket Working Group
(TBWG)
What is the ITER Test Blanket Module Program?
• The ITER Test Program is managed by the ITER Test
Blanket Working Group (TBWG) with participants from
the ITER Central Team and representatives of the Parties
• Breeding Blankets will be tested in ITER, starting on Day
One, by inserting Test Blanket Modules (TBM) 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 nuclearelectromagnetic environment
• Each ITER Party is allocated limited space for testing two
TBM’s. (No. of Ports reduced to 3. Number of Parties
increased to 6)
• ITER’s construction plan includes specifications for TBM’s
because of impacts on space, vacuum vessel, remote
maintenance, ancillary equipment, safety, availability, etc.
Stages of FNT Testing in Fusion Facilities
Fusion
“Break-in”
Stage:
Required
Fluence
2
(MW-y/m )
Size of Test
Article
I
~ 0.3
SubModules
• Initial exploration of
performance in a fusion
environment
• Calibrate non-fusion tests
• Effects of rapid changes in
properties in early life
• Initial check of codes and data
• Develop experimental
techniques and test
instrumentation
Design Concept
& Performance
Verification
Component Engineering
Development &
Reliability Growth
II
III
1-3
>4-6
Modules
Modules
/ Sectors
• Tests for basic functions and
phenomena (tritium release / recovery,
etc.), interactions of materials,
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 )
• Data on initial failure modes and
effects
• Narrow material combination
and design concepts
• Establish engineering feasibility of
blankets (satisfy basic functions &
performance, 10 to 20% of lifetime)
• 10-20 test campaigns, each is 12 weeks
• Select 2 or 3 concepts for further
development
• Identify failure modes and effects
• Iterative design / test / fail / analyze /
improve programs aimed at
improving reliability and safety
• Failure rate data: Develop a data
base sufficient to predict mean-timebetween-failure with sufficient
confidence
• 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
D
E
M
O
FNT Requirements for Major Parameters for Testing in Fusion Facilities with
Emphasis on Testing Needs to Construct DEMO Blanket
- These requirements have been extensively studied over the past 20 years, and they have been agreed to internationally
(FINESSE, ITER Blanket Testing Working Group, IEA-VNS, etc.)
- Many Journal Papers have been published (>35)
- Below is the Table from the IEA-VNS Study Paper (Fusion Technology, Vol. 29, Jan 96)
Parameter
a
Neutron wall load (MW/m2)
Plasma mode of operation
Minimum COT (periods with 100% availability) (weeks)
Neutron fluence at test module (MW·y/m2)
Stage I: initial fusion break-in
Stage II: concept performance verification (engineering feasibility)
c
Stage III : component engineering development and reliability growth
Total neutron fluence for test device (MW·y/m2)
Total test area (m2)
Total test volume (m3)
Magnetic field strength (T)
Value
1 to 2
b
Steady State
1 to 2
0.3
1 to 3
c
4 to 6
>6
>10
>5
>4
a - Prototypical surface heat flux (exposure of first wall to plasma is critical)
b - If steady state is unattainable, the alternative is long plasma burn with plasma duty cycle >80%
c - Note that the fluence is not an accumulated fluence on “the same test article”; rather it is derived from testing
“time” on “successive” test articles dictated by “reliability growth” requirements
Key Fusion Environmental Conditions for Testing Fusion
Nuclear Components
Neutrons (fluence, spectrum, spatial and temporal gradients)
-
Radiation Effects
(at relevant temperatures, stresses, loading conditions)
Bulk Heating
Tritium Production
Activation
Heat Sources (magnitude, gradient)
-
Bulk (from neutrons)
Surface
Particle Flux (energy and density, gradients)
Magnetic Field (3-component with gradients)
-
Steady Field
Time-Varying Field
Mechanical Forces
-
Normal
Off-Normal
-
Combined environmental loading conditions
-
Interactions among physical elements of components
Thermal/Chemical/Mechanical/Electrical/Magnetic Interactions
Synergistic Effects
R&D Tasks to be Accomplished Prior to Demo
1) Plasma
- Confinement/Burn
- Disruption Control
- Current Drive/Steady State
- Edge Control
2) Plasma Support Systems
- Superconducting Magnets
- Fueling
- Heating
3) Fusion Nuclear Technology Components and Materials
[Blanket, First Wall, High Performance Divertors, rf Launchers]
- Materials combination selection and configuration optimization
- Performance verification and concept validation
- Show that the fuel cycle can be closed (tritium self-sufficiency)
- Failure modes and effects
- Remote maintenance demonstration
- Reliability growth
- Component lifetime
4) Systems Integration
Where Will These Tasks be Done?!
• Burning Plasma Facility (ITER) and other plasma devices will address 1, 2, & much of 4
• Fusion Nuclear Technology (FNT) components and materials require dedicated fusion
facility(ies) parallel to ITER (prior to DEMO) in addition to TBM testing in ITER.
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.
Excellent opportunities exist for collaboration
between US and Korea on fusion engineering
• US has extensive experience in fusion blanket
systems developed over 30 years
• US has focused blanket R&D on key areas of
blanket feasibility
• Korea has strong background in fission and now
fusion technology systems
• Korea has strong industrial and manufacturing
capabilities
• Collaboration possibilities are numerous,
especially on development and deployment of
ITER TBMs of joint interest.
Possibilities for US-Korea Collaboration on
Helium-Cooled Ceramic Breeder Blankets
• Development and characterisation of ceramic
breeder and beryllium pebbles
• Thermo-mechanics of pebble beds
• Tritium release characteristics of ceramic
breeders and beryllium
• Beryllium behaviour under irradiation
• Helium cooling technology
• Prototypical mock-up testing in out-of-pile
facility
• In-pile testing of sub-modules
• Development of instrumentation
Utilization of Fission Reactors for Assessing Blanket
Feasibility Issues (examples)
Pebble bed assemblies for
thermomechanical
experiments at HFR in Petten
Tritium release experiments
in JMTR
Feature: a stepping motor used
in a fission capsule test to
50
cm
Module 1
Li4SiO4
T= 650 oC
simulate ITER pulsed operations
f60
Stepping motor
Hollow cylinder for
neutron absorber(Hf)
Li2TiO3
f65
pebbles f20
Aluminium(Al)
Hollowcylinder
(Hf)
Fixedneutron
absorber (Hf)
Windowangle
B
C
AandCsections
65 65 65
260
A
Windowof Hf neutronabsorber
Core
f65
Opencondition
Submodule
Module 3
Li2TiO3
T= 650 oC
B section
A : Cross section A
B : Cross section B
C : Cross section C
: Hot junction point of multi-paired thermocouple
: Self powered neutron detector (SPND)
Core
10.8 cm
Module 2
Li4SiO4
T= 850 oC
Closecondition
Module 4
Li2TiO3
T= 850 oC
Possibilities for US-Korea Collaboration 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
*Similar possibilities exist also for molten-salt blankets