This presentation is a brief glance about material challenges in fusion technology including an introduction to nuclear fusion principal, inertial confinement fusion, magnetic confinement fusion, materials for first-wall in tokamak principal, types of breeders, magnetohydrodynamic problem.

1. Karim Hossny
E-Mail: Hossny.K@Gmail.com
Alexandria University
Faculty of Engineering
Nuclear & Radiation Department
Associate Prof. Mohammed Hassan
E-Mail: MHMHEg@Yahoo.com
6/19/2014

2. Team Leader: Karim Hossny
E-Mail: Hossny.K@Gmail.com
Phone no: +2 0106 93 80 868
Team Members:
1. Abd El-Rahman Magdi
2. Akram Said Farag
3. Remon Samir
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3. 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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Different Fusion Fuel Scenarios

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Vacuum Pumping Duct
Diverter Plates
FW/Blanket
Vacuum Vessel
Shield
Toroidal Field
Coil
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Magnetic Confinement Fusion Reactor

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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

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Materials Arrangement in a Tokamak

8. 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
– high sputtering but less
quenching.
 High Z-W, Mo based alloys
– low sputtering but high
quenching.
First Wall
Heat Sink
 Cu-Cr-Zr alloy
 Copper alloys – dispersion
strengthened by Alumina.
First Wall
Structural
 Steels
 Vanadium alloys.
 SiC-fiber/SiC composites.
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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
 𝐿𝑖6
enriched liquid lithium or eutectic Pb-
17at%Li.
 𝐿𝑖6
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, 𝐸𝑟2 𝑂3 or 𝑌2 𝑂3.

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(n,2n) Cross-section of Pb-208, Jendl 6

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Absorption Cross-section of Li-6, JENDL 4

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Elastic, Gamma Production Cross-section of Li-7, JENDL 4

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Table 3 Concepts of Solid TBMs Proposed by Various Partners of ITER
Design
Parameters
China Europe Japan Korea Russia USA India
Option HCCB HCCB HCCB HCCB HCCB HCCB HCCB
Breeder
𝐿𝑖4 𝑆𝑖𝑂4
(400
− 950 ℃)
𝐿𝑖4 𝑆𝑖𝑂4
(450
− 900 ℃)
𝐿𝑖2 𝑇𝑖𝑂3
(900 ℃)
𝐿𝑖4 𝑆𝑖𝑂4
(400
− 900 ℃)
𝐿𝑖4 𝑆𝑖𝑂4
(1000 ℃)
Not
Decided
𝐿𝑖2 𝑇𝑖𝑂3
(850 ℃)
Neutron
Multiplier
Be (400 −
620 ℃)
Be (450 −
600 ℃)
𝐵𝑒/𝐵𝑒12 𝑇𝑖
(600 ℃)
Be (450 −
600 ℃)
Be (650 ℃) Be (500 ℃) 𝐵𝑒/𝐵𝑒12 𝑇𝑖
(600 ℃)
Structure
Eurofer
(530 ℃)
Eurofer
(550 ℃)
F82H Eurofer FMS (600 ℃) FMS
(550 ℃)
LAFMS
Coolant
He
(300 −
He
(350 −
Water (150-
250) bar
He
(350 −
He
(300 −
He
(300 −
He
(300 −
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

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Table 4 Concepts of Liquid TBMs Proposed by Various Partners of ITER
Design
Parameters
China Europe Korea Russia USA India
Breeder
and Coolant
Pb-Li (480 −
700 ℃)
He cooled
(DFLL)
Pb-Li (530 ℃)
He cooled
(HCLL)
Li (530 ℃)
He cooled
Li (350 −
550 ℃)
Li cooled
Pb-Li (500 ℃)
He cooled
(DCLL)
𝐿𝑖2 𝑇𝑖𝑂3
ceramic and
Pb-Li eutectic
Pb-Li liquid
cooled (LLCB)
Neutron
Multiplier
Be (550 ℃)
Structure CLAM(530 ℃)
Eurofer
(550 ℃)
Eurofer
(550 ℃)
V alloy FMS Indian LAFMS
Electro-
insulator
𝑆𝑖𝐶𝑓/𝑆𝑖𝐶
𝐴𝑙2 𝑂3
SiC
CaO, AIN,
𝐸𝑟2 𝑂3, Yttria
𝑆𝑖𝐶𝑓/𝑆𝑖𝐶
Flow Channel
Inserts
𝐴𝑙2 𝑂3
Reflector Graphite
WC/TiC
(600 ℃)
SS 316 SS 316 L

17. 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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18. The PbLi liquid-metal enters the blanket
modules at 460°C and leaves at 650°C to
700°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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19. 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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20. 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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