Download - Material Challenges in Fusion Technology
Karim Hossny
E-Mail: [email protected]
Alexandria University
Faculty of Engineering
Nuclear & Radiation Department
Associate Prof. Mohammed Hassan
E-Mail: [email protected]
6/19/2014
Team Leader: Karim Hossny
E-Mail: [email protected]
Phone no: +2 0106 93 80 868
Team Members:
1. Abd El-Rahman Magdi
2. Akram Said Farag
3. Remon Samir
6/19/2014Material Challenges in Fusion Technology 2
1.Introduction for Fusion Technology.
2.Materials for Tokamak.
3.ITER.
4.TBMs Materials.
5.Li-Self Cooled TBM.
6.Dual Coolant TBM.
7.MHD Coating.
8.MHD Coating Requirements.
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Material Challenges in Fusion Technology 4 6/19/2014
Different Fusion Fuel Scenarios
Material Challenges in Fusion Technology 5
Vacuum Pumping Duct
Diverter Plates
FW/Blanket
Vacuum Vessel
Shield
Toroidal Field
Coil
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Magnetic Confinement Fusion Reactor
Material Challenges in Fusion Technology 6
Bottom Blanket
Module
Bottom Access Flange
(Non-Breeding)
First Wall
Laser Beam
Shield
Blanket Support
Stud
First Wall Upper Access Flange
(Non-Breeding)
Upper Blanket
Module
Chamber
Support Column
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Inertial Confinement
Fusion Reactor
Material Challenges in Fusion Technology 7 6/19/2014
Materials Arrangement in a Tokamak
Plasma Outboard
Top
Bottom
Inboard
Table 1 Materials for the First Wall of a
Tokamak
First Wall
Plasma
Facing
Low Z-Be, C-C composites
β high sputtering but less
quenching.
High Z-W, Mo based alloys
β low sputtering but high
quenching.
First Wall
Heat Sink
Cu-Cr-Zr alloy
Copper alloys β dispersion
strengthened by Alumina.
First Wall
Structural
Steels
Vanadium alloys.
SiC-fiber/SiC composites.
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6/19/2014Material Challenges in Fusion Technology 9
ITER is the first magnetic confinement fusion
experimental reactor, designed to test
material in the true fusion environment in
order to make sure of its capability for future
commercial fusion reactors.
For testing such materials there must be
testing modules compatible with the testing
port in ITER.
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Testing modules mainly are designed to test
Tritium Breeding in addition to some effect of
radiation on some structural material.
USA does not have a Testing module to be
tested in ITER, instead they are designing
Fusion Nuclear Science Facility to test their
own blankets.
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Table 2 Functional Materials in TBMs
For Neutron
Multiplication
Beryllium, Be-8at%Ti (beryllide), BeO in solid
form.
Liquid lead
For Tritium Breeding
πΏπ6 enriched liquid lithium or eutectic Pb-
17at%Li.
πΏπ6 enriched ceramics like lithium titanate and
lithium silicate.
For Tritium Extraction He (purge gas through the ceramic breeder)
Liquid lead lithium eutectic.
For Self-Heeling
Coatings
Alumina on FMS.
AIN, CaO, πΈπ2π3 or π2π3.
Material Challenges in Fusion Technology 12 6/19/2014
(n,2n) Cross-section of Pb-208, Jendl 6
Material Challenges in Fusion Technology 13 6/19/2014
Absorption Cross-section of Li-6, JENDL 4
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Elastic, Gamma Production Cross-section of Li-7, JENDL 4
Material Challenges in Fusion Technology 15 6/19/2014
Table 3 Concepts of Solid TBMs Proposed by Various Partners of ITER
Design
ParametersChina Europe Japan Korea Russia USA India
Option HCCB HCCB HCCB HCCB HCCB HCCB HCCB
Breeder
πΏπ4πππ4(400
β 950β)
πΏπ4πππ4(450
β 900β)
πΏπ2πππ3(900 β)
πΏπ4πππ4(400
β 900β)
πΏπ4πππ4(1000β)
Not
Decided
πΏπ2πππ3(850 β)
Neutron
Multiplier
Be (400 β
620β)
Be (450 β
600β)
π΅π/π΅π12ππ
(600 β)
Be (450 β
600β)
Be (650 β) Be (500 β) π΅π/π΅π12ππ
(600 β)
StructureEurofer
(530 β)
Eurofer
(550 β)
F82H Eurofer FMS (600 β) FMS
(550 β)
LAFMS
Coolant
He
(300 β
He
(350 β
Water (150-
250) bar
He
(350 β
He
(300 β
He
(300 β
He
(300 β
Purge Gas He 0.5 bar He 0.5 bar He 0.5 bar He 0.5 bar He 0.5 bar He 0.5 bar He 0.5 bar
Material Challenges in Fusion Technology 16 6/19/2014
Table 4 Concepts of Liquid TBMs Proposed by Various Partners of ITER
Design
ParametersChina Europe Korea Russia USA India
Breeder
and Coolant
Pb-Li (480 β
700β)
He cooled
(DFLL)
Pb-Li (530 β)
He cooled
(HCLL)
Li (530β)
He cooled
Li (350 β
550β)
Li cooled
Pb-Li (500 β)
He cooled
(DCLL)
πΏπ2πππ3ceramic and
Pb-Li eutectic
Pb-Li liquid
cooled (LLCB)
Neutron
MultiplierBe (550 β)
Structure CLAM(530 β)Eurofer
(550 β)
Eurofer
(550 β)V alloy FMS Indian LAFMS
Electro-
insulator
πππΆπ/πππΆ
π΄π2π3SiC
CaO, AIN,
πΈπ2π3, Yttria
πππΆπ/πππΆ
Flow Channel
Inserts
π΄π2π3
Reflector GraphiteWC/TiC
(600 β)SS 316 SS 316 L
The common advantages of liquid Li cooled
concepts originate from the characteristics of
pure Li such as high thermal conductivity, high
heat capacity, high Li atomic density and low
tritium pressure due to its the high solubility
of tritium.
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The PbLi liquid-metal enters the blanket
modules at 460Β°C and leaves at 650Β°C to
700Β°C. The performed MHD calculations show
that the pressure drop in the PbLi channels of
the blanket due to magnetic/electric
resistance is small, if all walls are covered by
a SiC electric insulation of 5 mm thickness.
When projected for a reference tokamak
power reactor design, it has the potential for
a gross thermal efficiency of > 40%.
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MHD pressure drop and MHD flow
control are critical and common issues
for liquid metal self-cooled blanket
concept.
For self-cooled blanket concepts, MHD
insulators will be needed to reduce the
MHD pressure drop with a reduction
factor in the range of 10 to 100.
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1.High electrical resistivity.
2.Chemical stability with lithium.
3.Ability of coating complex channel
configurations.
4.Irradiation resistivity.
5.Self-healing of any defects occurring.
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