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. ALEXANDRIA UNIVERSITY
FACULTY OF ENGINEEIRNG
NUCLEAR & RADIATION ENGINEERNG DEPARTMENT
Introduction to Fusion
Technology Issues
Material Challenges
Associate Professor Mohammed Hassan
E-Mail: MHMHEg@Yahoo.com
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
3. Remon Samir
5/31/2014
This report is developed to give a brief introduction to fusion technology then marching to material
requirements for fusion reactors passing through different proposed blanket module designs.

2. May 31, 2014 [INTRODUCTION TO FUSION TECHNOLOGY ISSUES]
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Table of Contents
Introduction to Fusion Technology...............................................................................................................2
Magnetic Confinement Fusion..................................................................................................................2
Inertial Confinement Fusion .....................................................................................................................4
Material Issues in Fusion Reactors................................................................................................................5
Introduction ..............................................................................................................................................5
Materials for Tokamak..............................................................................................................................5
First Wall Materials...................................................................................................................................7
Materials for Other Components of TBM.................................................................................................8
ITER Test Blanket Module Functional Materials.........................................................................................10
Liquid Breeder TBM Concepts ................................................................................................................10
Self-Cooled Breeder Designs...................................................................................................................11
Li-Breeder Self-Cooled Designs...........................................................................................................11
Dual Coolant Designs ..........................................................................................................................11
MHD Coating Design Requirements ...................................................................................................11
Conclusion...................................................................................................................................................13
References ..................................................................................................................................................14
Consulted References .............................................................................................................................14

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Introduction to Fusion Technology
Fusion technology is like a new method of obtaining energy from nuclear reactions only this time
it’s not due to absorption of neutron but it is due to fusing two nuclei together depending on the
mass defect between them to get the energetic fusion products (Depending on the two fusing
nuclei) and 14.1 MEv neutron (Which is the main material concern due to its high energy compared
with fission neutron 2MEv).
The Physics problem of fusion is due to the requirement of fusing two positive nuclei together
overcoming the repulsion force between them, that’s why special environments are required for
achieving such fusion reaction.
There are two famous methods for achieving fusion reaction between deuterium and tritium atoms
which are the Magnetic Confinement Fusion and the Inertial Confinement Fusion, choosing the
two atoms to be fused together depends on the energy needed for achieving the reaction and cross-
section of the reaction itself, that’s why D-T fusion reaction is the most preferable reaction (see
figure 1).
Figure 1 Fusion Fuel Cycles
Magnetic Confinement Fusion
Magnetic confinement fusion of D-T depends mainly on pressure and temperature produced from
confined plasma (which is the environment upon which the fusion occur). Fusion produces 14.1
MEv neutron and alpha particles (collected in the divertor) (see figure 2).
+ → (3.5 ) + (14.1 )

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Figure 2 Magnetic Confinement Fusion
Plasma’s role is to create the required energy and pressure for both ions (D and T) in order to
overcome the in-between positive repulsive magnetic fields created by each ion.

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Inertial Confinement Fusion
Inertial confinement fusion depends on highly energetic laser beams (directed to the fuel pellet) to
overcome the positive magnetic repulsion force between the two ions (D-T), after that once the
fusion process has occurred it starts to burn out (from inside to outside) in an explosion producing
14.1MEv neutron and alpha particles (See figure 3)
+ → (3.5 ) + (14.1 )
Figure 3 Inertial Confinement Fusion

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Material Issues in Fusion Reactors
Introduction
Because of the 14.1 MeV neutrons that are generated in the + reaction exploited in a tokamak,
the materials, especially those employed for the construction of the first wall, the diverter and the
blanket segments, suffer crippling damage due to the high / ratios that result due to the
high energy of the neutrons. To meet this challenge, the materials that need to be developed for
the tokamaks are steels for the first wall and other structurals, copper alloys for the heat
sink, and beryllium for facing the plasma. For the TBMs, the materials that need to be developed
include beryllium and/or beryllium-titanium alloys for neutron multiplication, lithium-bearing
compounds for tritium generation, and the liquid metal coolants like lead-lithium eutectic in
which lead acts as a neutron multiplier and lithium as a tritium breeder. The other materials that
need attention of the materials scientists include superconductors made of NbTi, and
for the tokamaks, coatings or ceramic inserts to offset the effect of corrosion and the
MHD in liquid metal cooled TBMs, and a host of other materials like nano-structured materials,
special adhesives and numerous other alloys and compounds. Apart from this, the construction
of the tokamaks would necessitate development of methodologies of joining the selected materials.
Materials for Tokamak
Revising figure 2, the plasma is confined by the torodial and polodial magnetic fields in the form
of a ring in a vacuum vessel that has the shape of a toroid and the heat from the fusion in the plasma
is extracted by an appropriate coolant, the He gas and/or a eutectic alloy liquid flowing in the
blanket modules in the vacuum vessel close to plasma. The heat is transported to the coolant
through the walls of the TBMs by both radiation from plasma and the electrically neutral 14.1
neutrons that escape from the plasma into their walls and the functional materials.
Since the blanket consists of either in the form of a ceramic compound or liquid metal (pure
lithium or lead-lithium eutectic alloy), it transmutes to tritium by ( , ) reaction giving rise to
additional heat to the coolant. Further, when the 14.1 neutron escaping from the plasma
enter the walls of the TBM, complications arise both due to the radiation damage
(displacements and transmutations) of lattice atoms caused by them.
Because of the high cross section of these high energy neutrons to cause the ( , ) and the ( , )
reactions with almost all elements, atoms constituting the walls of the TBMs undergo these
reactions leading to the formation of both helium and hydrogen in them at high rates causing
serious damage to the structural material.

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Figure 4 A Schematic View for the Arrangement of Materials in Tokamak
The material behavior at the high / ratios (dpa, displacements per atom is the unit in which
the displacement damage of the lattice is expressed) likely to be encountered by the materials of
the first wall of the tokamak as well as the materials in the TBMs is yet not completely
understood. The challenge to put appropriate structural and functional materials in a tokamak as
well as in a blanket module in a configuration to serve the purpose desired from these devices for
the intended time is, indeed a challenge for the materials scientists. When the design and
construction of the TBMs for even the experimental ITER is considered, the relevance of the points
put forward until now becomes further evident.
The first wall of the Tokamak is the wall that is nearest to the plasma and, therefore,
experiences, the high / ratios due to the damage due to the high energy neutrons apart from
the high heat flux. The diverter and the limiter also fall in the same category. If material sputters
into the plasma, it may get quenched. To avoid this from happening, an element that either does
not sputter due to the neutrons (and, occasionally, electrons and other ions from the plasma) hitting
it or, else, it does not quench the plasma despite the fact that it sputters is selected. High Z (atomic
number) elements fall in the first category in that they sputter less and the low Z elements, even
though they may sputter into the plasma, they are not strong enough to quench it. The selection
of the plasma facing element is based on this. Once selected, this element has to be an integral
part of the first wall.

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Next to it in the first wall, especially in the diverter, has to be a material that can act as a heat sink
and carrier of heat away from the first wall to avoid its excessive heating. This generally is OFHC
(oxygen free high conductivity copper) alloyed with a little bit of (< 1 %) to give the Cu
the required tensile strength) and even a lesser content of (< 0.1 %) to impart the required
fatigue strength. Next to the listed plasma facing material or directly bonded to it, is the structural
material, generally a steel. This is the one that actually takes the entire load. Initially, austenitic
stainless steel 316 was selected for use as the first wall structural and continues to be material of
construction for the first wall of ITER in the form of low activation 316 LN (IG), IG meaning
the ITER grade. However, because of its tendency to swell more under irradiation as compared
to the ferritic steels and unacceptable fatigue life above 600℃, especially with He (generated due
to (n,α) reactions) in it, the material of choice for the first wall now for the DEMO reactors is the
low activation Ferritic/Martensitic (F/M) steel (FMS), F82H, or, its equivalents.
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
 Low activation austenitic steels [SS 316L(N) IG] for
the first wall.
 Ferritic/martensitic steels (F82H, EUROFER) for the
TBMs.
 Nanostructured ferritic/martensitic ODS steels or
nanostructured high nitrogen carbide dispersion
strengthened (CDS) F/M steels for the PROTOTYPE.
 Vanadium alloys.
 SiC-fiber/SiC composites.
First Wall Materials
Table 2 Comparison between the Properties of Various Structural Materials
Property
Material
FMS V-4Cr-4Ti SiCf/SiC
Temperature Window,
℃
300 − 600 400 − 700 700 − 1000
Surface Heat Capability,
/ .
4.32 − 2.74 4.61 − 4.63 1.05
Thermal Expansion,
/
11.1 − 12.3 10.3 − 11.4 2.5
Thermal Conductivity,
/ .
33.4 − 32.3 31.3 − 33.8 12.5
DBTT, ℃ < 20 250 − 300

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The critical issues related to the first wall materials include their transmutation and
displacement damage due to the high-energy neutrons, manufacturing the large sized intricate
shapes and their joining and codes for qualification of the materials for use in fusion
environments. So far as the damage due to neutrons is concerned, all the effects that occur in the
core of fast reactors occur in the fusion environment also, but more intensively. Helium produced
because of the ( , ) reactions of the neutrons with the atoms constituting the first wall is an issue
that is difficult to deal with. The rate of production of He in the material due to its irradiation
particularly by the 14.1 neutrons in a tokamak is very high (in the range of 200-600
appm/yr for steel) and, therefore, in its lifetime of 30 years, the material is likely to accumulate
huge amounts of He. Since the solubility of He in any metallic matrix is known to be zero, the high
temperature helium embrittlement is an issue of major concern. Furthermore, this He, under
thermal fatigue likely to be experienced by the first wall of a tokamak, limits the life of the first
wall austenitic steel severely. To overcome this challenge, the F/M steel has been substituted for
the stainless steel 316 as this has a much better thermal conductivity. This is being further
tackeled by distributing He into nano-sized bubbles by developing ODS F/M steel of 3rd
generation in which yttria particles having sizes less than 3nm diameter are distributed in large
numbers (10 / ). Further, the nanosized (18-20 nm dia) yittria gets refined to less
than 3nm dia during attrition of its mixture with steel powder only in the presence of Ti and,
therefore, this is to be added to the mixture before attrition. Ti-Y-O complexes form due to
attrition. Interestingly, Ti is the only element that can effectively achieve this. The reason is yet
to be established. Besides, the Ti-Y-O complexes act as sites for the nucleation of He bubbles.
The other issue relates to manufacturing of components, particularly joining of materials. Friction
stir welding, electro-discharge welding, and diffusion bonding by HIP are the technologies
that are currently being developed to advanced levels for meeting this challenge.
Materials for Other Components of TBM
It is seen that the TBM has to perform two main functions. It has to breed tritium (the naturally
non existing fuel for the fusion reactor), with a TBR more than one and also extract the heat
efficiently. Keeping these functions in view, a number of concepts have been proposed to
design the TBMs, first for the ITER. Some of these are termed as solid test blanket modules
and some as liquid test blanket modules, the difference being on the physical state in which the
breeder material is in the TBM. If the breeder (basically, ) is in the form of a solid ceramic
compound, it is solid breeder TBM and, if the breeder is in liquid state (as pure Li liquid or eutectic
Pb-Li alloy liquid), it is called a liquid breeder TBM. In the case of a solid TBM, the coolant, more
often than not, is He. In one such concept proposed by Japan, it is water. To have enough neutrons
for the breeding reaction, Be or beryllide is to be inserted in the solid TBM as a neutron multiplier.
The solid TBM thus consists of the structural material (low activation F/M steel), the ceramic
breeder (lithium titanate or lithium silicate), the neutron multiplier (Be or beryllide) and the
coolant, He. The material ofconstruction of TBM has been chosen to be F/M steel to gain
experience with this material as this is a candidate for the first wall of a DEMO.

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When Pb-Li is used, Li works as the breeder and Pb as the neutron multiplier. The liquid itself
sometimes is made to act as the coolant as well. As a coolant, it creates the extra issue of Magneto-
Hydro-Dynamic (MHD) drag on its own flow in the TBM, which raises further requirements in
terms of electrically insulating coatings on steel to reduce the drag, powerful pumps to push the
liquid through the TBM and, of course, the integrity of the material under forced flow at high
temperature of liquid metal. However, obviously, there is no need to insert Be or beryllide for
neutron multiplication in this case. The concepts of both the solid and liquid TBMs proposed by
the various partners in ITER.
Table 3 Functional Materials in TBMs
For Neutron Multiplication
 Beryllium, Be-8at%Ti (beryllide), BeO in solid form.
 Liquid lead
For Tritium Breeding
 enriched liquid lithium or eutectic Pb-17at%Li.
 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, or .
Table 4 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 (400
− 950 ℃)
(450
− 900 ℃)
(900 ℃) (400
− 900 ℃)
(1000 ℃)
Not
Decided (850 ℃)
Neutron
Multiplier
Be (400 −
620 ℃)
Be
(450
− 600 ℃)
/
(600 ℃)
Be
(450
− 600 ℃)
Be
(650 ℃)
Be
(500 ℃)
/
(600 ℃)
Structure
Eurofer
(530 ℃)
Eurofer
(550 ℃)
F82H Eurofer FMS
(600 ℃)
FMS
(550 ℃)
LAFMS
Coolant
He (300 −
500 ℃)
80 bar
He
(350
− 550 ℃)
80 bar
Water
(150-250)
bar
He
(350
− 500 ℃)
80 bar
He
(300
− 500 ℃)
80 bar
He
(300
− 550 ℃)
80 bar
He
(300
− 550 ℃)
80 bar
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 5 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)
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
/
SiC
CaO, AIN,
,
Yttria
/
Flow
Channel
Inserts
Reflector Graphite
WC/TiC
(600 ℃)
SS 316 SS 316 L
ITER Test Blanket Module Functional Materials
Liquid Breeder TBM Concepts
Liquid breeder TBM designs are proposed by different parties. The Russian Federation (RF) is
proposing testing of Li-self cooled TBM with Be as the neutron multiplier to enhance the tritium
breeding, and vanadium alloys as the structural material. Japan is considering the installation of
liquid breeder TBMs such as Li-self cooled TBM without Be, or FLiBe-self cooled TBM in the
later period of the ITER operation and testing. The European Union (EU) is focusing on the
helium-cooled PbLi concept (HCLL), where helium is used as the primary coolant to extract the
blanket power. For higher thermal performance the US is proposing to test a dual coolant PbLi
breeder concept (DCLL), where helium is used to cool all RAFMS structures, and the self-cooled
breeder is circulating slowly in order to reach a high exit temperature. This concept is also
proposed as a blanket option for the EU Power Plant Conceptual Study. China is proposing to test
blanket concepts called dual coolant PbLi (DLL) and single coolant PbLi (SLL) designs, which
are similar to the DCLL and HCLL concepts, respectively. For the DC designs FCIs are required
as thermal and MHD insulators to separate the high temperature PbLi from the lower temperature
RAFMS structures. To avoid the MHD issue of the self-cooled concept, Korea is proposing a He-
cooled blanket with quasistagnant liquid Li as the breeding material (HCML). Its thermal
performance is limited by the use of RAFMS.

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Self-Cooled Breeder Designs
For self-cooled breeder concepts there are two basic approaches. The first one is to use liquid Li
to perform the tritium breeding and heat removal functions. The second one is the DC concept
where helium is used to cool all the RAFMS structures and PbLi is the self-cooled liquid breeder.
MHD pressure drop and MHD flow control are critical and common issues for liquid metal self-
cooled blanket concept. The use of MHD insulator barriers to decouple electrically the flowing
liquid metal and the wall are necessary to reduce the pressure drop in order to control the system
pressure to an acceptable level. In general, 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.
Li-Breeder Self-Cooled Designs
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. V-alloys such as V-4Cr-4Ti were used as structural
material which has a maximum design limit of 700°C. A thermal efficiency of ~40% is projected
for the tokamak power reactor design. MHD coatings or FCIs are applied to the internal wall of all
Li flowing channels.
Dual Coolant Designs
The DC designs being proposed by the EU, US and China use high pressure helium to cool the
RAFMS structure and PbLi as the self-cooled breeder. The basic approach of the DCLL concept
shows the use of helium to cool the first wall and all RAFMS structural elements, and the use of
FCI elements to perform the key functions of reducing the MHD effect of the circulating PbLi.
FCIs made of SiC composite material in the PbLi channels serve as thermal and electric insulators
to minimize the MHD pressure loss and reach high coolant exit temperature and, thus, a high
efficiency of the power conversion system. 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%.
MHD Coating Design Requirements
For MHD coating, a thin ceramics coating on the inner surfaces of the channel wall has been
proposed. The principal requirement for the coating, in addition to resistivity, is compatibility
between the flowing liquid metal and the substrate wall materials. In the case of liquid Li-self
cooled blanket with vanadium structures, the highly reducing environment of Li narrows the option
of candidate ceramics.

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The requirements for the MHD insulator coating for Li/V blanket are:
1. High electrical resistivity, within acceptable property change in the operating environment
including radiation effects.
2. Chemical stability and compatibility with Li to the maximum operation temperature.
3. Mechanical integrity and thermal expansion match with V-alloy.
4. Safety/environmental characteristics, e.g. low activation.
5. Potential for coating on complex channel configurations.
6. Irradiation resistant.
7. In situ self-healing of any defects that might occur.

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Conclusion
Nuclear fusion is another way of getting energy from nuclear interaction, except this time its
between two nuclei not a nucleus and a particle. Energy from nuclear fusion (17.5 MeV) is way
less than energy released per fission (200 MeV). On the other hand fusion is cleaner energy than
that from fission as there are no fission fragments, only neutrons are produced which can be easily
shielded by water (which also act as a part from the heat removal system).
Since the most famous or “Easy” fusion process is the D-T one, and D exists in see water, the
problem exists in T which up till now ITER will depend on Tritium produced from CANDU
reactors; which has lead them to try to transmutate Lithium into Tritium and that’s the point of
testing different breeding blanket designs.
ITER is a research reactor which will have port in which different modules will be tested in order
to have the most optimum blanket design which will achieve highest TBR in order to achieve
sustainability of future reactors.
Future D-T fusion reactors will be Tritium self-sufficient by having their blankets made of one of
the ITER tested blanket modules, the point is optimizing the radiation effect, TBR and the heat
removal system.
Regarding other future fusion reactors depending in D-D fuel pellets there won’t be fuel problem,
but on the other hand there would be a problem achieving the fusion itself.

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References
1. Materials Issues in Fusion Reactors, A K Suri, N Krishnamurthy and I S Batra, Materials
Group, Bhabha Atomic Research Centre, Journal of Physics: Conference Series 208
(2010), 23rd
National Symposium on Plasma & Technology (PLASMA-2008), IOP
Publishing.
2. ITER Test Blanket Module Functional Materials, C.P.C. Wong, V. Chernov, A. Kimura,
Y. Katoh, N. Morley, T.Muroga, K.W. Song, Y.C. Wu and M. Zmitko, General Atomics,
November 2005.
Consulted References
1. The Challenge of Developing Structural Materials for Fusion Power Systems, Everett E.
Bloom, Oak Ridge National Laboratory, Metals and Ceramics Division, Journal of Nuclear
Materials 258-263 (1998).
2. Current Status of Fusion Reactor Structural Materials R&D, Akira Kohyama, Institute of
Advanced Energy, Kyoto University, Materials Transactions, Vol. 46, No. 3 (2005),
Special Issue on Fusion Blanket Structural Materials R&D in Japan.
3. HCCB Summary Supplements, Alice Ying, August 2005.
4. Introduction to Fusion Technology Issues, Lecture II, In Vessel Components: Blanket,
Shield Divertor, Mohamed Sawan, Fusion Technology Institute, University of Wisconsin-
Madison, September 2013.