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Nuclear fusion as a massive, clean, and inexhaustible energy source for the second half of the century: Brief history, status, and perspective


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Fusion energy, based on the use of broadly available inexhaustible resources as lithium and deuterium and with minimal impact to the environment, aims at a change in the energy supply paradigm: instead of its current dependence on natural resources and environmental impact, energy would become a technology-dependent resource with unlimited adaptive availability and whose unit cost should decrease as technology progresses. This article intends to give a picture of where fusion research stands today and the perspectives: the achievements, the difficulties, the current status, marked by the construction of the ITER experiment which will demonstrate the scientific feasibility of fusion power, and the perspectives toward the first demonstration power plant, DEMO, which, according to the European Roadmap, could start the construction shortly after the full power experiments in ITER (<2030) and be in full operation, generating net electricity into the grid, by 2050.
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Nuclear fusion as a massive, clean, and inexhaustible
energy source for the second half of the century: brief
history, status, and perspective
Joaquin S
Laboratorio Nacional de Fusion, CIEMAT, 28040 Madrid, Spain
Fusion, Energy, Plasma
J. S
anchez, Laboratorio Nacional de Fusion,
CIEMAT, 28040 Madrid, Spain.
Tel: +34913466159; Fax: +34913466124;
Funding Information
No funding information provided.
Received: 14 April 2014; Revised: 16 July
2014; Accepted: 17 July 2014
doi: 10.1002/ese3.43
Fusion energy, based on the use of broadly available inexhaustible resources as
lithium and deuterium and with minimal impact to the environment, aims at a
change in the energy supply paradigm: instead of its current dependence on
natural resources and environmental impact, energy would become a technol-
ogy-dependent resource with unlimited adaptive availability and whose unit
cost should decrease as technology progresses. This article intends to give a pic-
ture of where fusion research stands today and the perspectives: the achieve-
ments, the difficulties, the current status, marked by the construction of the
ITER experiment which will demonstrate the scientific feasibility of fusion
power, and the perspectives toward the first demonstration power plant,
DEMO, which, according to the European Roadmap, could start the construc-
tion shortly after the full power experiments in ITER (<2030) and be in full
operation, generating net electricity into the grid, by 2050.
It is widely recognized that the energy supply is one of
the largest challenges that mankind will be facing during
this century. Population growth and increasing per capita
consumption of goods and services in the emerging coun-
tries will lead to a likely twofold energy demand in a cou-
ple of decades, despite the efforts toward efficiency and
energy savings in the developed countries. In addition,
new demands will appear derived from the need for mas-
sive water supply and food production or large-scale recy-
cling of basic materials.
In this scenario, we will need to count on massive
sources of energy, environmentally friendly, and based on
abundant primary resources. Nuclear fusion intends to be
one of these sources, its main objective being to trans-
form the energy paradigm: from today’s dependence on
natural resources and environmental impact into a tech-
nology-dependent resource, in the same way as we see
today Internet access, mobile communications, or com-
puter power: a resource whose availability can grow easily
with demand and whose cost per unit decreases as tech-
nology progresses.
Nuclear fission, the basic process in today’s nuclear
power plants, consists on breaking a large nucleus into
medium size ones, nuclear fusion is based on the opposite
reaction: the union of two small nuclei in order to gener-
ate a larger one, but still small. In both cases, the mass of
the reaction products is slightly smaller than that of the
original nuclei and this lost mass is converted into energy
according to Einstein’s equation E = mc
. The nuclear
forces involved in the process are much larger than the
electromagnetic forces which are the basis of standard
combustion of fuel and so is the capability of energy pro-
duction per unit mass of fuel: one gram of fusion fuel,
equivalent to about 7 tons of oil [1] would be enough to
provide the full energy consumption of an average person
during more than 1 year.
Fusion is the reaction which powers the sun and all the
stars. In the center of a star, the incredible high gravita-
tional forces generate the conditions for the fusion of
hydrogen nuclei into deuterium as the first of a chain of
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and John Wiley & Sons Ltd.
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reactions in which deuterium will fuse with remaining
hydrogen into helium (He
) and later the helium with
itself (to generate He
). On Earth, where we cannot count
on such strong gravitational forces, we will need to look
for more accessible reactions, though still very hard to
achieve. The fusion reaction with larger cross section
under reasonably achievable conditions is the fusion of
deuterium and tritium, two isotopes of hydrogen. The
reaction produces as a result a helium nucleus and a neu-
tron and releases 17.6 MeV (mega electron volt) of
energy, (91,000 kWh per gram of fuel). Of this energy, 4/
5 is carried by the neutron and the remaining 1/5 by the
helium nucleus.
A fusion power plant would be essentially a thermal
plant. The energy released by the fusion reaction is
absorbed by a coolant and extracted to the heat exchang-
ers and to the electricity-producing turbines. The fusion
fuel would be composed of two species: deuterium and
tritium. Deuterium exists in natural water in a fraction of
33 mg/L. On the other hand, tritium, another, heavier,
hydrogen isotope, is unstable and does not exist in nat-
ure. It is usually a secondary product of fission power
plants. Fusion reactors would generate in situ the tritium
they would consume by means of neutron bombardment
of lithium, another chemical element. Lithium is also very
abundant in nature and, given the fact that the required
quantities are very small in comparison with the amount
of energy obtained, it could be extracted at affordable cost
from salts solved in seawater. The estimated reserves of
lithium in seawater would be sufficient to satisfy the
world’s energy needs during many million years and it is
expected that in the future technologies for mastering the
deuteriumdeuterium reaction would become available,
thus extending the availability of fusion fuel beyond the
expected life of the solar system.
The main exhaust product resulting from the reaction
is helium, the very same gas we use to fill balloons for
children. This element is harmless for people and the
environment, it does not contribute to the greenhouse
effect and, in fact it does not even accumulate in the
atmosphere: due to its low weight it escapes to the space.
In addition, the quantities produced would be very small:
if all the energy in the world would come from fusion,
the amount of helium produced worldwide would be in
the order of several thousand tons a year, to be compared
with the ten billion ton CO
per year released currently.
Safety is a major concern on every industrial facility
and in nuclear plants in particular. One of the advantages
of a nuclear fusion plant would be its intrinsic safety:
fusion plants will be safe not just because they will be
carefully designed and operated, they will be safe because
the physical properties of the process make impossible an
uncontrolled fusion reaction. As we will discuss later, the
very high temperatures required in the reactor, in the
order of several hundred million degrees, are impossible
to sustain in case any malfunction arises, for example, an
air leak into the reactor would immediately bring down
the temperature and extinguish the fusion reaction.
Another element to be taken into account is that, whereas
in a fission reactor the amount of fuel inside the reactor
could sustain the reaction for months (and therefore it
might be more difficult to manage if control is lost), in a
fusion plant the fuel contained inside the reactor would
last for only a few seconds if the supply from outside is
interrupted. As an example: the cooling system, whose
failure was the cause of the problems at the Fukushima
reactor, was not even an important safety component in
the large experiment ITER because there was no signifi-
cant residual heat when the operation stops.
The main safety concern in a fusion plant would be the
existence of tritium, which is radioactive and, even if it is
not a long-term pollutant (its half life is 12.3 years), it is
dangerous if inhaled or ingested as tritiated water. Fortu-
nately, tritium is only used as a transition element and
the main supply comes as lithium, but still the storage of
several kilograms of tritium is difficult to avoid and, as it
would happen with any other dangerous substance in an
industrial facility, safety measures are required to prevent
any release to the environment. The current designs
would guarantee that in the worst case accident, there
would be no need to evacuate people staying outside the
facility fence.
The main drawback of fusion as a potential source of
energy is the difficulty to generate and sustain the reac-
tion. In order to achieve the reaction, the two colliding
nuclei must get close enough to allow the short range
nuclear forces to act, this can only be achieved if their
energy is high enough to overcome the electrostatic repul-
sion of the two positively charged nuclei. Accelerating
deuterium or tritium ions to these energies, 1520 keV, is
not particularly difficult, the difficulty arises when we try
to get energy gain from the process: launching an ion
beam against a target at the required energy would pro-
duce just a very small fraction of fusion reactions because
in most cases the long range coulombian repulsion will
deviate the ions, which will miss the target, and also
many of them will release their energy in collisions with
the electrons. The only way to achieve an efficient process
is to be able to confine the accelerated ions in a closed
space, in such a way that, after having gained the required
energy, they have many opportunities to collide before
their energy is lost. The main problem is the availability
of a suitable recipient: a gas where the average particle
energy is 15 keV has a temperature of 170 million degree.
Since 1950s, scientists have been trying to find this
kind of recipient and have developed two main families
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Nuclear Fusion J. S
of experiments: inertial confinement, based on a fast heat-
ing of the fuel so that it enters the fusion reaction before
it has time to expand, and magnetic confinement, based
on the fact that at such high temperatures the gas, in state
of “plasma”, is composed of charged particles, which can
be confined by magnetic fields.
Inertial fusion uses an ion beam or a laser, the pre-
ferred option nowadays, as the means of delivering a big
amount of energy in a very short time to the deuterium
tritium (DT) target. The target is illuminated with spheri-
cal symmetry and this produces a pressure wave which
converges toward its center. At a given moment, a fusion
“spark” should be generated in the center and the heat
generated by these initial fusion reactions would propa-
gate back the fuel burn toward the rest of the target. The
most advanced inertial fusion experiment is currently the
National Ignition Facility (NIF) located in the Lawrence
Livermore National Laboratory, (Livermore, CA).
Recently, experiments have been reported where the ini-
tial spark of fusion has been found [2]. Its extension to
the whole target has not yet been achieved and the energy
production is still a small fraction of that delivered by the
lasers, however it is a promising result. A similar experi-
ment, the “Laser M
egajoule” is under construction in
France, essentially with military purposes: inertial fusion
experiments can be used to validate the models which are
the basis for the computer simulation of thermonuclear
In parallel, the largest worldwide effort toward fusion
energy has been and is being devoted to the so called
“magnetic confinement”. The DT fuel at such high ener-
gies is on “plasma” state, a gas where ions and electrons
move separately, and can, therefore, be confined by a
magnetic field, which essentially allows particles to move
freely along the field lines but forces them to move in
small circles when trying to go in perpendicular direction.
The next step is to construct a configuration where field
lines close on themselves, so, ideally, particles would stay
indefinitely moving along those closed trajectories. These
configurations have been implemented since the 1960’s
along two main families of toroidally shaped devices:
“stellarators” and “tokamaks” which differ in the way
they generate the necessary complementary field which is
required to avoid particle drift in a toroidal geometry.
The third approach, a linear configuration with “magnetic
mirrors” at both ends, turned to be much less effective
and has been less developed.
Fusion Devices: The “Tokamak” and
The “tokamak” word derived from the Russian expres-
sion for “toroidal chamber with magnetic coils” was
first developed by I. Tamm and A. Sakharov (who later
received the Nobel peace prize) in the early 1960’s and
was rapidly adopted by researchers around the world.
Thirty years later (1991), the Joint European Torus (JET)
a tokamak experiment owned by the European Com-
mission and located in Culham, near Oxford (UK) car-
ried out the first D-T experiment toward controlled
fusion, providing a substantial amount of energy from
the fusion reactions [3], few years later (199697), JET
and the TFTR tokamak (Princeton, NJ) reached fusion
power levels in the order of 1015 MW, with a ratio of
fusion power to heating power of 60% [4, 5].
Despite the criticism to fusion researchers to be “always
40 years away” from the goal of fusion power, the reality
is that the efficiency of tokamak experiments, measured
as the “triple product” of ion temperature, ion density,
end energy confinement time (T
), was growing at
comparable pace to that of microprocessors between 1960
and 2000 and will hopefully recover when the large ITER
experiment will start (see Fig. 1). However, it is necessary
to realize the magnitude of the challenge: the magnetic
field confines very well a single particle, but, as the many
particles collide, there is diffusion across field lines and
both particle and energy flow slowly away. In order to
minimize these losses we have two essential tools: one is
to increase the magnetic field, but this has a limit for the
superconductor coils which generate it, so it is difficult to
envisage a device with an average field above 67 Tesla;
the second tool is to increase the machine size. “Wind
tunnel” comparisons with tokamaks of similar geometry
and increasing size have shown that in order to achieve
“ignition” conditions, a situation where the energy gener-
ated in the fusion reactor can compensate the losses and
maintain the required high temperatures which sustain
the reaction, we need a hot plasma volume in the order
of 1000 m
for a standard magnetic field value of 56T.
This means very large, complex, and expensive devices,
with development times in the range of 1020 years.
After the success of JET and TFTR, the next step will
be ITER, a joint experiment of seven parties which repre-
sents more than half the world population: China, India,
Japan, Korea, Russia, the United States, and Europe,
which acts as a single party and is represented by the
European Commission. ITER (from the latin word "iter",
the way), with nearly 1000 m
of hot plasma, will aim at
demonstrating the scientific feasibility of fusion as an
energy source. The specific objective is to obtain energy
gain Q = 10, which means that ITER will generate
~500 MW of fusion power with 50 MW of external
power being injected to heat the plasma. The gain Q = 10
would be sustained during 400 sec periods; as a second
objective, a less demanding value of Q = 5 would be sus-
tained for periods of 1500 sec (see details in Table 1) In
ª 2014 CIEMAT, Spanish Ministry of Economy and Competitiveness. Energy Science & Engineering published by the Society of Chemical Industry
and John Wiley & Sons Ltd.
J. S
anchez Nuclear Fusion
addition, ITER will carry out a number of experiments to
test the technology developments necessary for a power
plant, in particular, the “breeding blanket” modules
which will test the technology for tritium generation from
lithium (Fig. 2).
ITER is a large extrapolation in volume (10 times lar-
ger than JET) and also in technology. In addition to the
use superconducting coils, cooled at 1.4 K while located
less than a meter away from the million degree hot
plasma, the largest challenges come from the goal to
operate long pulses at full fusion power: all the internal
elements need active cooling as well as neutron-resistant
functional materials, particularly insulators.
The challenge in science and technology is formidable
but it is not the only one: ITER is also to some extent a
social experiment. With magnetic fusion research being a
declassified activity both by the eastern and western coun-
tries during the cold war, the undertaking of a large joint
experiment was one of the agreements between presidents
Reagan and Gorbachev on their summit of November
1985 at Geneva. The European Union and Japan joined
the project immediately and the design evolved slowly
during the following years in a process which included
the temporary withdrawal of the USA and the decision in
1998 to redesign the device in order to make it more
affordable and ended with the delivery of the final design
report in 2001. In 2004, the USA returned to the project
and China and Korea, as well as later India, expressed
their interest to join, which was welcome in order to
share the multibillion costs of the project.
The main drive for the interest of the parties was fusion
energy as a long-term goal, but, in the shorter term, their
interest was also focused in the important technology
developments around ITER. This led to an organization
based on a moderate size central team, located at the pro-
ject site in Cadarache (France) and in charge of about 15%
of the total ITER budget, and seven, smaller but still strong,
teams, named “domestic agencies”, located at the different
parties’ headquarters and in charge of delivering in-kind
components to the central team for about 85% of the bud-
get. Europe as the host party would provide nearly half of
the total budget and the remaining part would be covered
by the other six parties.
Table 1. Main ITER parameters.
Fusion power 500 MW
Fusion power gain (Q) >10 (for 400 sec inductively driven
>5 (1500 sec)
Plasma major radius (R) 6.2 m
Plasma minor radius (a) 2.0 m
Plasma vertical elongation 1.70/1.85
Plasma current (Ip) 15 MA
Toroidal Field at 6.2 m
5.3 T
Installed auxiliary heating 73 MW
Plasma volume 830 m
Plasma surface area 680 m
Plasma cross-section area 22 m
Figure 1. Progress of the fusion triple product T
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Nuclear Fusion J. S
The organization based on in-kind contributions
allowed for an a priori distribution of the participation
and could accommodate the wish of the parties to partici-
pate in the technologies of their interest, in addition, this
was the way to allow for the emerging countries to have
lower costs by developing components with their own
workforce. On the other hand, the system, based on a
central team which prescribes the design but does not
have responsibility on the cost of construction of these
components and the domestic agencies which have to
procure and pay for the components, is prone to produce
internal discussions and delays in the decisions.
The ITER agreement was signed 21 years after the idea
was launched, in November 2006 and the first estimate of
the construction period was 9 years. Soon it became evi-
dent that between the report delivered in 2001 and the nec-
essary constructive design there was much more distance
than originally estimated. The report had concentrated in
the main machine parameters, the related physics and the
design of the critical high-technology components, but
ITER was a very complex industrial plant, subject to a
nuclear license, not as a nuclear power plant but as a
nuclear facility, and with a very demanding integration
process into the buildings’ design. The consequence of hav-
ing concentrated on the critical components, necessary to
guarantee the feasibility of the project, but having over-
looked the more conventional parts of the facility was an
underestimation of cost, which essentially doubled after an
in depth revision, and construction time.
Although the cost has been kept within reasonable
bounds after the 2010 revision, suffering moderate
increases but remaining within the limits of the originally
foreseen contingency, the schedule seems difficult to con-
trol. Subsequent revisions of the baseline schedule have
led to an estimate for the “first plasma”, which would
mark the end of the construction period, to happen in
202223. This delay is the accumulation of several causes:
lack of a finalized design, lack of manpower at the central
team imposed by the budget restrictions, delays in Japa-
nese components after the 2011 earthquake damaged
some key facilities, additional licensing requirements
derived from the postFukushima revision of all the
nuclear procedures, etc., but a significant part of the delay
comes from the extremely complex organization of the
project and the distribution of roles and responsibilities.
A typical example is when a component design performed
by the central team is felt as an over specification which
rises cost by the domestic agency in charge of the con-
struction: the domestic engineers will come back with
redesigns aiming to lower the cost while the central ones
will be just worried by confidence in the functional role
of the component, thus entering on a loop with no clear
outcome. Many of those organizational problems have
been highlighted by the management assessment report
commissioned recently by the ITER Council.
In the meantime, the good news is that most of the
high-tech components like the superconducting coils or
the vacuum vessel, have undergone the final designs, with
the corresponding design reviews, and the related con-
struction contracts have been awarded to industry, which
is so far progressing without known major difficulties.
The major technical problem happened with the central
solenoid superconducting cable, which in 2012 was show-
ing degradation with operation time in the samples
tested. Fortunately, further R&D by the Japanese team in
charge of this cable provided, on time for the coils con-
struction at the USA, a new design which was successfully
tested without showing any degradation.
Figure 2. Artist view of the ITER device.
ª 2014 CIEMAT, Spanish Ministry of Economy and Competitiveness. Energy Science & Engineering published by the Society of Chemical Industry
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J. S
anchez Nuclear Fusion
Other elements of confidence have been provided from
the physics side by the research carried in supporting exper-
iments around the world. As an example, one of the
elements of concern with the original design of ITER,
which used a carbon inner wall, was the problem of tritium
accumulation in the form of hydrocarbons deposited in
remote parts of the device, which could lead to the require-
ment to stop operation after every few experiments and
undertake a complex tritium removal procedure. On the
other hand, the use of carbon, due to its good behavior at
high temperature and low atomic number, was capital for
an efficient operation from the physics point of view and
no clear alternative was at sight.Fortunately,testsofplasma
operation with a full tungsten wall carried out in the recent
years in the German device ASDEX-U and with a tungsten
divertor and beryllium wall (the same combination of
ITER) in JET have demonstrated reliable efficient operation
without the tritium retention problem. Now ITER has
changed its design and the lower part of the inner wall, the
so called “divertor”, where the interaction with the plasma
concentrates, will use tungsten as plasma-facing material.
In parallel, developments in the control of the periodic
busts of power to the wall (the so called Edge Localised
Modes, ELMs), the progress in the understanding of
energy and particle confinement and its extrapolation to
ITER size or the achievement of reliable operation at the
high plasma densities projected for ITER reinforce our
confidence in the operational success of the experiment.
Still some of the original concerns remain, for instance
the need to avoid and mitigate the so called “disrup-
tions”, rapid losses of confinement which could lead to
damage of internal components, but progress is steady in
all those fronts.
This situation, with organizational delays in one side
but smooth progress in the most critical components, and
physics projections in the other side, makes us to be rela-
tively optimistic toward the actual success of the project
and encourages us to work in order to find the right
organizational frame to avoid further delays.
With ITER starting operation in 2023, the critical high-
gain results with Q = 10 will come shortly before 2030.
One of the answers we expect to get from these experi-
ments is the efficiency of the plasma heating by the high-
energy He ions generated at the fusion reaction, also
called “alpha particles”. This is crucial for the future of
the tokamak as a fusion reactor because we need to use
these alpha particles to maintain the high temperatures
which sustain the reaction. In the fusion reaction, neu-
trons carry 80% of the released energy, they cannot be
retained by the magnetic field, and therefore they cannot
contribute to sustain the reaction (in the power plant
their energy will be extracted by the coolant and used to
drive the turbines). Alphas will carry only 20% of the
fusion power but they are charged particles which can be
retained in the plasma by the magnetic field and contrib-
ute to sustain the plasma temperature. The problem is
that, whereas the plasma particles have an average energy
of 1530 keV, the alphas are born with an energy of
3.5 MeV, hundred times higher, and they would escape
quickly unless the energy transfer mechanism by means of
collisions is efficient enough. Preliminary experiments in
JET [6] as well as theoretical predictions show that, very
likely, the alphas will indeed heat the plasma efficiently,
but the ultimate test will be performed in ITER. With
Q = 10, the power generated by fusion will be ten times
the heating power injected externally, then the alphas,
which carry 20% of this power will provide twice the
externally injected heating, leading to a clear effect that
will serve as a concluding test of the alpha heating.
Toward the Demonstration Reactor
ITER will demonstrate the scientific feasibility of fusion
as energy source and will also test key technologies for
the reactor but ITER will not yet be a real power plant.
The main differences between ITER and a demonstration
power plant, the so called “DEMO”, from “demonstra-
tion”, would be: tritium self-sufficiency, full plant energy
efficiency, use of low-activation neutron-resistant materi-
als, and reliable continuous operation. In the following
pages, we will address the status and perspective of the
related developments.
Tritium Self-Sufficiency
As explained in the introduction, fusion plants would
need to generate in situ the tritium they will consume by
bombarding lithium with the fusion-generated neutrons,
through the reaction: n+Li
. This function will
be performed by the so called “breeding blanket” which
will surround the plasma. The breeding blanket has also
two main additional functions: to extract the power of
the neutrons conveying it to the steam generators and the
turbines and to shield the sensitive components, in partic-
ular the superconducting coils, from the neutron flux
which would heat and damage them. This makes of the
breeding blanket a very demanding nuclear component.
ITER will not be equipped with a full breeding blanket,
and therefore it will not be self-sufficient in tritium. The
current plan is to purchase the tritium that it will con-
sume from an external source, essentially the Canadian
nuclear fission programe, but it will have a number of
smaller blanket modules which will be used to test differ-
ent tritium-generation technologies.
The ITER blanket modules will test different options
for the three main elements in a breeding blanket. First,
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Nuclear Fusion J. S
there is the choice of breeding material, the main options
being molten eutectic lithiumlead, with 90% Li
ment or lithium salt pebbles (Li
or Li
) with
3060% Li
enrichment. Secondly, we need a neutron
multiplier, usually beryllium or the lithiumlead itself,
because a fraction of the neutrons generated in the fusion
reaction will fail to hit the breeding material. Fortunately,
the neutron energy required for the breeding reaction is
relatively low and using a neutron multiplier each single
14 MeV fusion neutron can generate several secondary
neutrons able to produce tritium. The third element is
the coolant, which must extract the energy deposited and
generated in the blanket (the breeding reaction is exother-
mic), here the options are: water cooling, helium cooling,
or dual cooling by helium and lithiumlead. [7].
The integral test of breeding blanket modules in ITER
will be a crucial experiment in order to validate the dif-
ferent technologies. The strategic value of those designs is
such that the breeding blanket program is not part of the
ITER agreement, which foresees that all the knowledge
generated in the project will be shared among the seven
parties, but a separate activity whose results would be pri-
vate intellectual property. It has to be coordinated due to
the host role of ITER but the information obtained in the
experiments will be sole property of the related party and
in principle would not be shared.
The European Roadmap toward fusion electricity [8]
includes a breeding blanket technology programe parallel
to the preparation of the validation tests of the ITER
blanket modules. The four technologies selected are the
two which Europe will test in ITER, lithiumlead, and
ceramic pebbles both cooled by helium, plus two addi-
tional options. The water cooled lithiumlead, as a
shorter term option, has the advantage of avoiding the
use of helium, which might become scarce if thousands
of fusion plants need to use it, and the high cooling
capacity of water; on the other hand, water generates cor-
rosion problems as well as safety issues and its tempera-
ture operation window (280325°C) is hardly compatible
with the low-activation neutron-resistant materials which
we have at hand today. The dual coolant, a longer term
option uses a faster circulation of the liquid LiPb to use
it as high-temperature coolant. The use of insulating
inserts and an additional helium cooling system allow for
the structural material to remain at lower temperature
than the main coolant, which has a much higher temper-
ature and leads to a higher plant efficiency.
High Gain for Plant Efficiency and the
Energy Extraction Problem
The Q = 10 power gain of the ITER plasma will not be
enough for having a real energy gain in the full balance
of the plant, which must take into account all the energy
consumption of the coils, cryogenic systems, and other
auxiliary systems as well as the wall plug efficiency of the
plasma heating systems and the efficiency of the thermal
cycle. Overall, an efficient power station would require Q
in the order of 50, which means a device with either a
more efficient physics, a higher magnetic field (difficult to
achieve due to the limitations in the superconductors) or
a larger size.
Typical European designs of a demonstration fusion
power plant [9] consider the total fusion power in the
range of 2000 MW thermal and have a linear size, 1.5
times of ITER. A device with this size and power is a sig-
nificant challenge, in particular, on what concerns the
extraction of the power.
As explained before, the neutrons carry 80% of the
power, they escape the plasma isotropically and cross the
wall of the tokamak and are absorbed volumetrically in
the coolant and blanket structures. The thermal power is
very high but this broadly distributed load can be toler-
ated by the materials. The neutrons can generate a num-
ber of other issues in the materials but the thermal load
is not a serious problem.
On the other hand, the remaining 20% of the power,
carried by the alpha particles, together with externally
injected power, also carried by charged particles, flows
slowly toward the wall. The magnetic field can delay this
flow but once the steady state is reached there is a contin-
uous flux of energy toward the inner wall of the device.
All this power is conveyed to a small fraction of the
wall, the so called “divertor”. This is necessary to avoid
the penetration of sputtered wall particles into the hot
plasma that would quench the high temperature but it
generates a serious problem: all this power is deposited in
a narrow, several cm, ribbon along the torus, leading to
thermal loads in excess of 20 MW/m
, twice higher than
the current engineering limit.
The possible solutions to what the experts see nowa-
days as the main challenge toward the success of fusion
energy, operate from the two sides of the problem: cool-
ing the plasma edge by emission of radiation in the visi-
ble and ultraviolet range, which distributes the load over
a larger surface, and designing “divertor” geometries and
materials which can handle the power.
Cooling of the plasma edge can be achieved by inject-
ing gases like nitrogen, krypton, and argon [10], the goal
is that a large fraction of the power is radiated at the edge
while preserving the good core confinement. In addition,
geometries which expand the interaction area in order to
decrease the power density are being developed, like the
“super X” [11] or the “snowflake” [12] divertors. The
final tool to overcome this challenge is the choice of
materials, the basic reference material is tungsten but
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J. S
anchez Nuclear Fusion
liquid metal alternatives, using lithium, gallium, or tin,
which offer “self-repairing” walls, are also being consid-
Neutron Resistant Materials
The power carried by the neutrons is not a big problem
as explained above, however, the high fluence of ener-
getic 14 MeV neutrons generates a different set of prob-
lems in the material. The problems will not be present
in ITER, at least for the structural materials, due to the
relatively low accumulated neutron fluence, but will be
very severe for DEMO and for the commercial fusion
Firstly, each neutron impact will give rise to a cascade
of collisions which will displace many atoms from their
positions. This is measured on “displacements per atom”
or dpa’s, one dpa meaning that, on average, every atom
within the material has been displaced once. The struc-
tural material of the blanket and first wall in a fusion
reactor will suffer an excess of 100 dpa’s during the com-
ponent lifetime, in addition, the 14 MeV neutrons, dis-
tinctly to the neutrons in a standard fission reactor, will
generate transmutation reactions in the material which
will produce helium and hydrogen and create blisters as
well as material swelling. All these phenomena can
degrade significantly the mechanical properties of the
material, but there is one more adverse effect: the irradi-
ated material becomes radioactive and will have to be
treated as radioactive waste.
The materials which adapt best to the 14 MeV neutron
bombardment are: vanadium alloys, titanium alloys, sili-
con carbide, a long-term promise but still difficult to use
as structural material and, the current reference material
which has achieved the highest technological maturity,
the RAFM (Reduced Activation Ferritic-Martensitic)
steels. As iron is relatively resilient to neutron bombard-
ment and suffers little activation, RAFM steels, like the
Japanese F82H or the European EUROFER, are based on
the suppression of problematic impurities (Ni, Cu, Al, Si,
Co, etc.) and the substitution of problematic alloying
components (Mo, Nb) by other elements which play the
same chemical role in the alloy but have a more benign
nuclear behavior (Ta, W). RAFM steels would suffer less
activation than the standard ones although they would
still be an activated material after decommissioned from
the fusion reactor. The current studies foresee that the
components could be recycled after ~100 years under cus-
tody as medium-low level radioactive waste, as opposed
to ~100,000 years for standard steel components under
equivalent conditions. The possibility to further reduce
this period depends on the level of impurity suppression
technically, and economically, achievable. A fast activation
decay is also observed for vanadium alloys [13] but vana-
dium currently lacks industrial development and has
some negative effects, like corrosion, Tritium permeation,
and narrower operating temperature.
One of the problems in the development of materials
for fusion reactors is the absence of intense sources of
14 MeV neutrons which could allow us to test the behav-
ior of the material in similar conditions to those in the
fusion plant [14]. EUROFER has shown good perfor-
mance under irradiation in fission reactors, which essen-
tially reproduce the dpa’s effect but there is little
knowledge about the effect of He and H accumulation.
One possibility is to use theoretical modeling of the
irradiation effects. Activation is relatively easy to deter-
mine, as it essentially depends on the concentration of
the different elements and neutron propagation calcula-
tions are possible. However, the structural changes are
nearly impossible to compute starting from first princi-
ples: we are in a problem where the number of particles
is in the order of Avogadro’s number and the changes
must be tracked in picoseconds scale for periods of many
seconds (which are the characteristic times of the changes
in the mechanical properties). The modeling is performed
using a multiscale approach, but the approximations are
such that the experimental tests of every scale model as
well as an overall test of the complete modeling are neces-
A family of 14 MeV neutron sources under consider-
ation is based on the use of reduced size fusion reactors
with modest Q but with substantial DT reaction rate sus-
tained by external injected power and equipped with a
full breeding blanket in order to self generate the tritium.
The so called CTF’s (Component Test Facilities) belong
to this family and there are several proposals under study
both in the China and the USA [1517].
The second family of sources is based on accelerator-
driven neutron generation. For example, the reference
proposal, IFMIF (International Fusion Materials Irradia-
tion Facility) considers two 40 MeV deuteron beams of
125 mA each which hit a liquid lithium target producing
a neutron spectrum very similar to that of a fusion reac-
tor. IFMIF, a 1500 M project, would produce 20
50 dpa/year in a reduced volume of 0.5 L and smaller
rates in the wider adjacent space. It is considered as the
ultimate tool to qualify materials for the fusion power
plants. Currently, Europe and Japan are carrying valida-
tion developments for IFMIF components and a complete
accelerator with all the basic elements will be tested in
Rokkasho (Japan) in 2017. The possibility to use this pro-
totype accelerator in an early reduced version of IFMIF is
currently gaining momentum. This source could be avail-
able by the early 2020’s in order to qualify components at
20 dpa for an earlier phase of DEMO.
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Nuclear Fusion J. S
In the meantime, the fusion materials programe is
strongly involved in the development of new materials
and the consolidation of the reference ones, for example,
one of the current limitations of EUROFER type steels is
their reduced operation temperature window (350
550°C) which might be expanded by using ODS (Oxide
Dispersion Strengthened) versions with yttrium oxide.
Limited irradiation tests are also carried in fission reactors
(use of boron doped material or the inclusion of some
amount of
Fe can simulate the He generation by
14 Mev neutrons) or using multiple ion beams to pro-
duce simultaneous dpa’s and He/H implantation, at a
very fast rate but in very reduced sample volumes. Those
experiments can complement the theoretical models as
Maintenance Issues and Plant
RAMI (Reliability, Availability, Maintainability, and In-
spectability) will be a key issue in a complex facility as a
fusion power plant if we want it to operate under eco-
nomically sustainable conditions. In particular, given the
fact that the structure will become activated soon after
the start of operation, most maintenance operations will
have to be done by remote control manipulation. This
means that all components inside the vacuum vessel and
many of the components inside the cryostat, even for
ITER, will have to be designed compatible with Remote
Handling (RH) operations: size and weight of the compo-
nents, assembly method, assembly sequence, interfaces
with the RH tools...etc. Today, devices like JET have
shown the feasibility of complex RH operations like the
full substitution of the divertor or the first wall (see
Fig. 3), however, the replacement times need to be signifi-
cantly shortened in a commercial reactor and this would
imply to evolve from today’s man-driven operation to
automatic operation for many of the actions.
A lot of technology development would also be
required in plant systems: tritium extraction, isotope sep-
aration systems, He, and liquid metal heat exchangers, as
well as advanced thermal cycles are among the systems
which are currently being developed as part of the fusion
technology programes worldwide.
Steady State Reactors: The
The “tokamak” concept, on which ITER and most fusion
devices are based, is a very clever design with optimal
confinement properties. In this configuration, the con-
fined plasma contributes itself to the construction of the
confining magnetic configuration, this is achieved by
inducing a strong electric current in the highly conductive
hot plasma. With this contribution from the plasma,
some complex additional magnets that otherwise would
be necessary are spared. The current also contributes to
heating the plasma by Joule effect. This solution offers
some advantages and disadvantages as compared with the
other family of devices, the “stellarator” which assumes
no help from the plasma and configures the complete
magnetic field by means of additional 3-dimensionally
shaped magnets.
The advantages of the tokamak: comparative simplicity
and very good confinement properties, makes this config-
uration the best option for a fusion ignition prototype
like ITER or even a first DEMO device, however, the
tokamak has also some limitations derived from the
strong coupling of the plasma and its confinement. First
of all, the plasma current (up to 15 MA on ITER) is usu-
ally induced with a transformer effect, which is impossible
to sustain in steady state. Today’s tokamaks are pulsed
devices and this might have implications on the manage-
ment of the supply to the electric network and the com-
ponents fatigue when used as a power plant. Some
progress has been achieved in the development of nonin-
ductive current drive systems, but there is still a long way
ahead in the path toward the complete steady state. The
second problem, derived from the plasmaconfinement
coupling, is the existence of scenarios where plasma and
confinement drop suddenly together in a very fast
positive feedback process (milliseconds) which ends in a
tremendous thermal release to the wall and the quench of
the >10 MA plasma current. In these events, called “dis-
ruptions”, very strong electromagnetic forces are
Figure 3. JET remote handling system.
ª 2014 CIEMAT, Spanish Ministry of Economy and Competitiveness. Energy Science & Engineering published by the Society of Chemical Industry
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J. S
anchez Nuclear Fusion
generated and jets of fast electrons can achieve multi
MeV energy, becoming a potential threat for the integrity
of the internal components.
On the other hand, Stellarators are inherently steady
state devices which could operate under stationary condi-
tions for months and stop only for maintenance pur-
poses. As the confinement is decoupled from the plasma,
stellarators are also free of disruptions.
Stellarators are, in fact, older than tokamaks, first
devices were developed by Spitzer in the early 1950
s, but
the simplicity and good results of the tokamak soon rele-
gated them to a secondary role. In the 1980
s, new design
tools and constructive techniques, together with the intro-
duction of radiofrequency plasma heating systems which
could substitute the traditional Joule heating based on
plasma current, allowed for a relaunching of the stellara-
tors and the results from devices like the German W7AS
and the Japanese LHD, a superconductor-based device,
have shown the strong potential of this configuration,
overcoming the main limitations in confinement that hin-
dered the progress with earlier devices. In 2015, the large
superconducting stellarator W7X (Fig. 4), currently under
construction in Greifswald (Germany), will start opera-
tion. The results of this experiment might strongly rein-
force the potential of the stellarator as a long-term option
for the commercial reactor units, on which the engineer-
ing complexity will play a secondary role compared with
the simplicity and smoothness of the operation.
Fission Fusion Hybrid Systems
The 14 MeV DT fusion neutrons can be used to irradiate
uranium 238 or thorium 232 and generate fissile material,
which could be used either in a pure fission reactor, in
this case the fusion system would be a way to produce fis-
sion fuel, or in the fusion reactor blanket playing the role
of energy amplification. The same DT neutrons could be
used to just irradiate and “burn” the nuclear radioactive
waste accumulated during the complete history of fission
energy generation.
Those three applications have intermittently gained and
lost attention since the idea was conceived in the 1950
and later relaunched by H. Bethe in 1979 [18]. In princi-
ple, a fusion gain Q = 5, complemented with a 109
amplification from the fission blanket would suffice for
having an efficient power plant, which means that from
the fusion side, a device like ITER, and even a bit smaller,
could do the job. Those who support the idea see as the
main advantage the simplification of the fusion core and
a faster process toward energy generation. For those
opposing, the hybrids just bring together all the problems
of fusion, in particular complexity, and fission: less waste
but still significant, proliferation, handling of highly active
material. A very interesting discussion, which includes the
opinion of a “skeptics group”, can be found in ref. [19].
Currently, there is no effort on hybrids in the European
Roadmap, which focus on “pure fusion”, but there are
active groups in China and the USA and significant activ-
ity and interest have been reported by the Russian prog-
rame [20].
The Roadmap to Fusion Electricity
The parallel effort of ITER and the technology programes
should converge in the construction of a DEMO power
reactor. The concept of DEMO varies in the different
Figure 4. Assembly process of the W7X stellarator.
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Nuclear Fusion J. S
world programes and it is not even clear whether DEMO
would be a single worldwide collaborative experiment like
ITER or several competing developments running in par-
allel in different countries, looking for a leading position
in a phase where the economic profit of fusion might be
at sight.
The European DEMO concept [21] sees the device as
the last experimental facility before industry takes the lead
in the construction of commercial fusion plants. As
described above, it should be self-sufficient in tritium, use
advanced low-activation materials and provide in the
order of 500 MW of net electricity to the grid during
operational periods of several weeks.
With the ITER high Q experiments foreseen for the late
2020’s and the results of the 20 dpa materials irradiation
available by the same dates, the DEMO construction
could start by the mid 2030’s and should be able to start
net electricity generation before 2050 (Fig. 5). By that
time, we expect that the materials irradiation facility IF-
MIF would have been built and provided the necessary
data for full qualification of low-activation structural
materials under > 100 dpa’s. Those data might also be
complemented with results from the current projects for
CTF. From this point, we will enter the situation where
private investors and industry will engage in the construc-
tion of the first commercial plants. When will this happen
2010 2020 2030 2040 2050
8. Stellarator rotaralletSamsalPgninruBnoitasimitporotaralletS
7. Low cost Low capital cost and long term technologies
5. Safety
4. Tritium breeding
Parallel Blanket Concepts
Fusion Engineering Testing Reactor (CFETR, CN) and Fusion Neutron Science facility (FNS, US)
3. Materials
2. Heat Exhaust
Baseline strategy
Advanced configuration and materials
Medium Sized Tokamaks, linear plasma devices and Divertor Tokamak Test Facility (DTT)
1. Plasma operation
Steady state regimes
Medium Sized Tokamaks
6. DEMO Component Design and Eng. Design Construction Operation
Early Neutron Source
Fusion Electricity
ITER steady
Figure 5. European Fusion roadmap [8].
ª 2014 CIEMAT, Spanish Ministry of Economy and Competitiveness. Energy Science & Engineering published by the Society of Chemical Industry
and John Wiley & Sons Ltd.
J. S
anchez Nuclear Fusion
is difficult to predict, it will depend on the energy market
situation and the overall energy supply scenario, but given
the size and potential profits of the energy market, (the
full cost of ITER construction, estimated 1215 billion
euro, is about the cost of one single day of worldwide
energy consumption) we expect that this might be a rela-
tively fast development, leading to a significant share of
fusion in the energy mix during the second half of the
Conflict of Interest
None declared.
1. Freidberg, J. P. 2007. P. 12 in Plasma physics and fusion
energy. Cambridge University Press, New York, NY.
2. Hurricane, O. A., D. A. Callahan, D. T. Casey, P. M.
Celliers, C. Cerjan, E. L. Dewald, et al. 2014. Fuel gain
exceeding unity in an inertially confined fusion implosion.
Nature 506:343347.
3. Rebut, P.-H, and the JET Team. 1992. The JET
preliminary tritium experiment. Plasma Phys. Control.
Fusion 34: 1749.
4. Keilhacker, M., A. Gibson, C. Gormezano, P. J. Lomas, P.
R. Thomas, M. L. Watkins, et al. 1999. High fusion
performance from deuterium-tritium plasmas in JET.
Nucl. Fusion 39:209234.
5. Strachan, J. D., S. Batha, M. Beer, M. G. Bell, R. E. Bell, A.
Belov, et al. 1997. TFTR DT experiments. Plasma Phys.
Controlled Fusion 39:B103B114.
6. Thomas, P. R., P. Andrew, B. Balet, D. Bartlett, J. Bull, B.
De Esch, et al. 1998. Observation of alpha heating in JET
DT plasmas. Phys. Rev. Lett. 80:55485551.
7. Giancarli, L. M., M. Abdou, D. J. Campbell, V. A.
Chuyanov, M. Y. Ahn, M. Enoeda, et al. 2012. Overview of
the ITER TBM program. Fusion Eng. Des. 87:395402.
8. Romanelli, F., P. Barabaschi, D. Borba, G. Federici, L.
Horton, R. Neu, et al. 2012. A roadmap to the realisation
of fusion energy. EFDA document available at http://www.
9. Maisonnier, D., I. Cook, P. Sardain, L. Boccaccini, L. Di
Pace, L. Giancarli, et al. 2006. DEMO and fusion power
plant conceptual studies in Europe. Fusion Eng. Des.
10. Kallenbach, A., M. Bernert, R. Dux, L. Casali, T. Eich, L.
Giannone, et al. 2013. Impurity seeding for tokamak
power exhaust: from present devices via ITER to DEMO.
Plasma Phys. Controlled Fusion 55, art. no. 124041
11. Kotschenreuther, M., P. Valanju, S. Mahajan, L. J. Zheng,
L. D. Pearlstein, R. H. Bulmer, et al. 2010. The super X
divertor (SXD) and a compact fusion neutron source
(CFNS). Nucl. Fusion 50, art. no. 035003
12. Ryutov, D. D., R. H. Cohen, T. D. Rognlien, and M. V.
Umansky. 2012. A snowflake divertor: a possible solution
to the power exhaust problem for tokamaks. Plasma Phys.
Controlled Fusion 54, art. no. 124050
13. Zucchetti, M., S. A. Bartenev, A. Ciampichetti, and R.
Forrest. 2007. A zero-waste option: recycling and clearance
of activated vanadium alloys. Nucl. Fusion 47:S477S479.
14. Zinkle, S. J., and A. M
oslang. 2013. Evaluation of
irradiation facility options for fusion materials research
and development. Fusion Eng. Des. 88:472482.
15. Wan, Y. 2012. Mission & readiness of a facility to bridge
from ITER to DEMO. 1st IAEA-DEMO Program
workshop, Los Angeles, USA, October 1518.
16. Peng, Y. K. M., J. M. Canik, S. J. Diem, S. L. Milora, J. M.
Park, A. C. Sontag, et al. 2011. Fusion nuclear science
facility (FNSF) before upgrade to component test facility
(CTF). Fusion Sci. Technol. 60:441448.
17. Wong, C. P. C., V. S. Chan, A. M. Garofalo, J. A. Leuer,
M. E. Sawan, J. P. Smith, et al. 2011. Fusion nuclear
science facility advanced tokamak option. Fusion Sci.
Technol. 60:449453.
18. Bethe, H. 1979. The fusion hybrid. Phys. Today 32:4451.
19. Freidberg, J.P., Fink, Ph, et al. 2009. Fusion-Fission
Research Workshop (2009) September 30October 2,
Gaithersburg, MD (USA). Available at
fusion-fission/ (accessed 04 07 2014).
20. Azizov, E. A., G. G. Gladush, and E. P. Velikhov. 2013.
Project development of experimental hybrid reactor on the
basis of a compact tokamak for the disposal of spent
nuclear fuel. Proceedings 21st International Conference on
Nuclear Engineering, July 29
August 2, Chengdou, China.
21. Federici, G., R. Kemp, D. Ward, C. Bachmann T.
Franke, S. Gonzales, C. Lowry, M. Gadomska, J.
Harman, B. Meszaros, C. Morlock, F. Romanelli, and R
Wenninger. 2014. Overview of EU DEMO design and
R&D activities, Fusion Engineering and Design 89:882
12 ª 2014 CIEMAT, Spanish Ministry of Economy and Competitiveness. Energy Science & Engineering published by the Society of Chemical Industry
and John Wiley & Sons Ltd.
Nuclear Fusion J. S
... The availability of fuel is the fusion reactor's first big benefit. It is, after all, a technologically reliant energy source that does not place pressure on any natural resource [75]. The main fuel of the fusion reactor is deuterium, a nonradioactive isotope of hydrogen. ...
... Presumably, this will be accomplished first by the industrialized nations, who are now leading the charge. Developing nations need not be in a rush in this instance; they can learn from the experience of the developed world, and if they find the initial cost-prohibitive, they can simply wait until setting up the reactors becomes more affordable, as economics will drive down costs and the cost per unit reactor will decrease [75]. ...
... Lack of pollution It does not produce any significant radioactive waste [75]. ...
... 5 More and more human and financial resources are being put into the fusion research. 6 As the largest international energy cooperation project including seven members, the International Thermonuclear Experimental Reactor (ITER) located at Cadarache in France is being constructed. It is expected to achieve the first plasma in December 2025 and deuterium-tritium (DT) operation in 2035. ...
... The third is tritium nonradioactivity loss, for example, part of tritium retained in the fusion materials and finally regarded as radioactive waste without recycle. The definition of tritium self-sufficiency and the required TBR was shown and calculated using the formula (6). ...
Full-text available
Commercial tritium resources available are too scarce to fully supply the future fusion reactors after International Thermonuclear Experimental Reactor (ITER). Tritium self‐sufficiency, ITER fails to fully validate, was regarded as one of the most important issues needed to be solved in the pathway of achieving fusion energy. After ITER, several concepts of fusion engineering test reactors and fusion demonstration reactors have been proposed worldwide, for example, Chinese Fusion Engineering Test Reactor (CFETR), Fusion Nuclear Science Facility (FNSF), DEMOnstration fusion reactor (DEMO) in European Union and Korea. CFETR is in the engineering design phase and would be hopefully completed around 2020. Tritium resources for the reactor start‐up and tritium self‐sufficiency are two primary issues besides the steady‐state operation for CFETR. The objectives of this work are as follows: (a) to introduce the preliminary fuel cycle concept and available tritium resources for CFETR, (b) to evaluate and discuss the tritium demand for CFETR start‐up (phase I: 200 MW) and the feasibility of DD start‐up, (c) to identify the possible pathways to tritium self‐sufficiency through sensitivity analysis based on the design baseline of CFETR, (d) to evaluate the consequences in case of failing tritium self‐sufficiency, and (e) to identify future R&D needed for tritium self‐sufficiency. It is expected to give insights into the question on how to start the reactor in a more economical way, into the feasibility of tritium self‐sufficiency, and into the question on what will happen in case of failing tritium self‐sufficiency.
... Nuclear fusion energy source has been considered as the most ideal and sustainable energy due to its numerous merits, such as clean, safety, high efficiency and abundant fuel resource [1,2]. As the key part of test blanket module (TBM) in fusion reactor, tritium breeding materials are used to produce the tritium during the fusion reaction process, and thus achieving tritium self-sufficiency [3,4]. ...
... The potential advantages can be concluded as follows: (1) As the grain size dramatically decreases, tritium diffusion distance within the grain is obviously shortened, which is helpful to improve the tritium release rate [13]. (2) Nanoceramics possess abundant grain boundaries, which not only provide more and faster tritium diffusion paths, but also act as pinning sites of radiation defects, thereby enhancing the irradiation resistance of tritium breeding materials [10]. (3) Due to the fine crystal reinforcing effect, the mechanical strength of the tritium breeding ceramics may also be enhanced [14]. ...
Li4SiO4 has been widely studied as attractive tritium breeding materials due to its innate merits. Considering the potential advantages of nanostructure in tritium breeding materials, a distinctive process was developed to obtain nanostructured Li4SiO4 pebbles. In brief, ultrafine precursor powders were synthesized by solvothermal method without using surfactants, and then indirect wet method was adopted to generate the green spheres with homogeneous microstructure. After that, the suitable sintering conditions were defined by studying the effects of sintering parameters on the grain size evolution, and nanostructured Ti-doped Li4SiO4 pebbles were first obtained by two-step sintering method. This study will be expected to provide references for fabricating other Li-based tritium breeding materials.
... Nuclear fusion energy is another type of nuclear energy that is under investigation. The advantages of the fusion reaction fuel are as follows [4,5]: (1) deuterium and tritium are the main fuels. Deuterium is plentiful in natural water. ...
Tritium is the key fuel in nuclear fusion reactors. With the development of the international thermonuclear experimental reactor (ITER) project, the annual requirement of tritium has increased up to several kilograms. The candidate materials for tritium storage have many shortcomings such as insufficient kinetic performance, disproportionation effect, poor oxidation resistance, and poor helium (He) retaining ability. Therefore, it is urgent to develop a novel material system which satisfies all the requirements of tritium storage materials. High-entropy alloys (HEAs) have a unique structure of severe lattice distortion and have attracted much attention as hydrogen storage materials due to their high storing capacity and great hydrogenation performance. The distorted lattice helps to provide more interstitial sites for accommodating H atoms and enhance the He retaining ability by slowing down the He diffusion in the HEA lattice. In this work, the current research status of tritium storage materials, including the background and the basic criterion of tritium storage materials, as well as the disadvantages of the current materials, has been reviewed. Moreover, the theoretical and experimental studies of HEAs, focusing on the hydrogenation properties and the defect evolution in the distorted lattice, have been summarized. The HEAs may have great potential as tritium storage materials due to their potential hydrogenation performance and He retaining ability. Finally, the existing challenges and future development directions are also proposed.
... Such devices are then intended to operate as scientific research facilities for several more decades. Fusion research and development for energy supply has thus far been based almost solely on a single technical approach: the tokamak (Ikeda, 2009, Sánchez, 2014. Tokamaks use magnetic fields to confine a plasma containing ionised isotopes of hydrogen, typically deuterium and tritium 1 , and auxiliary systems to heat the plasma to high temperature (~100 M Kelvin). ...
Full-text available
Despite several decades of dedicated R&D, fusion, a potentially world-changing energy source, remains decades away from commercialisation. The majority of development thus far has been via publicly-funded programmes led by government laboratories focused on scientific research and in which commercialisation strategy and innovation play a minor role. Generally, such programmes follow a linear model of innovation in which commercial aspects are not considered until later in development. In consequence and without intention, devices not well-suited for commercial application are being pursued. In recent years, however, privately funded fusion start-ups have emerged with the goal of accelerating the commercialisation of fusion. Fusion start-ups are, by necessity, operating on a fundamentally different model of innovation: agile innovation, whereby technology is developed flexibly and iteratively towards an explicit commercial goal. Technology Roadmapping is a method that has been effective for supporting agile innovation but thus far has had limited application to mission-led hardware development. We characterise the key features of the fusion innovation approach and create a novel Technology Roadmapping process for fusion start-ups, which is developed via a case study with Tokamak Energy Ltd. The main elements of the developed process, the resulting Technology Roadmap, and its impact are presented.
Hydrogen isotope storage materials are of great significance for controlled nuclear fusion, which is promising to provide unlimited clean and dense energy. Conventional storage materials of micrometer-sized polycrystalline ZrCo alloys prepared by the smelting method suffer from slow kinetics, pulverization, disproportionation, and poor cycling stability. Here, we synthesize a honeycomb-structured ZrCo composed of highly crystalline submicrometer ZrCo units using electrospray deposition and magnesiothermic reduction. Compared with conventional ones, honeycomb ZrCo does not require activation and exhibits more than 1 order of magnitude increase in kinetic property. Owing to low defects and low stress, the anti-disproportionation ability and cycling stability of honeycomb ZrCo are also obviously higher than those of conventional ZrCo. Moreover, the interfacial stress (due to hydrogenation/dehydrogenation) as a function of particle radius is established, quantitatively elucidating that small-sized ZrCo reduces stress and pulverization. This study points out a direction for the structural design of ZrCo alloy with high-performance hydrogen isotope storage.
Li4SiO4 has been regarded as the vigorous competitors for tritium breeding materials due to its distinct advantages. In this work, Fine-grained Li4SiO4 pebbles were prepared by a combined sol–gel and hydrothermal method (SG–H method). According to our researches, hydrothermal treatment was helpful to gain the ultrafine powders, and two-step sintering process showed obvious advantages in the preparation of fine-grained Li4SiO4 pebbles. The related physical parameters of the samples were measured. Afterwards, thermal cycling tests were carried out to study the microstructure evolution and the variation of crushing load. SEM analyses and compression tests results indicated the Li4SiO4 pebbles obtained by SG–H method exhibited favorable stability under thermal cycling. It suggested that the obtained Li4SiO4 pebbles might meet the requirement of high service temperature in fusion reactor blanket.
This work aimed to fabricate a gold coating on the surface of ultralow-density carbon-hydrogen foam cylinder without electroless plating. Poly (divinylbenzene/styrene) foam cylinder was synthetized by high internal phase emulsion, and chemical vapor deposition polymerization approach was used to form a compact poly-p-xylylene film on the foam cylinder. Conducting gold thin films were directly deposited onto the poly-p-xylylene-modified foam cylinder by magnetron sputtering, and electrochemical deposition was adopted to thicken the gold coatings. The micro-structures and morphologies of poly (divinylbenzene/styrene) foam cylinder and gold coating were observed by field-emission scanning electron microscopy. The gold coating content was investigated by energy-dispersive X-ray. The thicknesses of poly-p-xylylene coating and sputtered gold thin-film were approximately 500 and 100 nm, respectively. After electrochemical deposition, the thickness of gold coating increased to 522 nm, and the gold coating achieved a compact and uniform structure.
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A Fusion Nuclear Science Facility (FNSF) is necessary to make possible a DEMO of the Advanced Tokamak (AT) type after ITER. One candidate, Fusion Nuclear Science Facility-AT (FNSF-AT), should have neutron wall loading of 1-2 MW/m 2 , continuous operation for periods of up to two weeks, a duty factor goal of 0.3 on a year and neutron fluence of 3-6 MW-yr/m 2 in ten years to enable development of blankets suitable for tritium and electricity production while demonstrating nearly all the critical elements necessary for the qualification and design of a DEMO. FNSF-AT, also called FDF, will be designed using conservative implementations of all elements of AT physics to produce 150-300 MW fusion power with modest energy gain (Q<7) in a modest sized normal conducting coil device. It will demonstrate and its results will help in the selection of the DEMO tritium breeding blanket concept. It will demonstrate the tritium fuel cycle, the behavior of candidate plasma facing materials, and the design and cooling of the first wall chamber and divertor components. It will also provide experience in safe operation and remote maintenance necessary for the DEMO design.
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Ignition is needed to make fusion energy a viable alternative energy source, but has yet to be achieved. A key step on the way to ignition is to have the energy generated through fusion reactions in an inertially confined fusion plasma exceed the amount of energy deposited into the deuterium-tritium fusion fuel and hotspot during the implosion process, resulting in a fuel gain greater than unity. Here we report the achievement of fusion fuel gains exceeding unity on the US National Ignition Facility using a 'high-foot' implosion method, which is a manipulation of the laser pulse shape in a way that reduces instability in the implosion. These experiments show an order-of-magnitude improvement in yield performance over past deuterium-tritium implosion experiments. We also see a significant contribution to the yield from α-particle self-heating and evidence for the 'bootstrapping' required to accelerate the deuterium-tritium fusion burn to eventually 'run away' and ignite.
A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L-H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter P-sep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about P-rad,(main)/R-2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of P-sep/R have been achieved so far, and close to DEMO values of P-rad,P-main/R-2, albeit at lower P-sep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.
Successful development of fusion energy will require the design of high-performance structural materials that exhibit dimensional stability and good resistance to fusion neutron degradation of mechanical and physical properties. The high levels of gaseous (H, He) transmutation products associated with deuterium–tritium (D–T) fusion neutron transmutation reactions, along with displacement damage dose requirements up to 50–200 displacements per atom (dpa) for a fusion demonstration reactor (DEMO), pose an extraordinary challenge. One or more intense neutron source(s) are needed to address two complementary missions: (1) scientific investigations of radiation degradation phenomena and microstructural evolution under fusion-relevant irradiation conditions (to provide the foundation for designing improved radiation resistant materials), and (2) engineering database development for design and licensing of next-step fusion energy machines such as a fusion DEMO.
Conference Paper
The paper describes development of project of hybrid reactors fusion-fission based on compact tokamak. It can have several advantages: such tokamak because of the smallness of aspect ratio A has a smaller volume at the same power as compared with the classical tokamak. At the same time, in their design, one can use database of ITER. The paper presents the results of a long phase of work in this direction, directed to the installation of the industrial scale of interest for use in nuclear power. FNS-2 parameters are R = 2 m, a = 1 m, k = 1.7, B = 3.9 T. Calculations show at neutral injection power ∼ 45 MW, the neutron yield is ∼ 60 MW, it provides a neutron load on the blanket ∼ 0.4 MW/m2. Numerical calculations showed that channel structure blanket cooled by water will transform for the year to 80 kg of minor actinides at subject to the addition of Pu in a 1:1 ratio). When a liquid metal coolant is used, the design of the outer blanket FNS-2 is similar to the block structure of the ITER. During the year this hybrid reactor loading 60 tons of oxides of actinides can recycle 400 kg of minor actinides. This will help dispose of spent nuclear fuel more than 10 LWR-1000 reactors. In this case the thermal power is ∼ 2 GW. The pressure which provides lead pumping through a stainless steel pipe is 1.5 bar at the length of 20 cm.
One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100 MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.
The objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such information is essential to design and predict the performance of DEMO and future fusion reactors. It foresees to test six mock-ups of breeding blankets, called Test Blanket Module (TBM), in three dedicated ITER equatorial ports from the beginning of the ITER operation. The TBM and its associated ancillary systems, including cooling system and tritium extraction system, forms the Test Blanket System (TBS) that will be fully integrated in the ITER machine and buildings. This paper describes the main features of the six TBSs that are presently planned for installation and operation in ITER, the main interfaces with other ITER systems and the main aspects of the TBM Program management.
The Fusion Nuclear Science Facility (FNSF) aims to address Fusion Energy Sciences research needs in ``Materials in Fusion Environment''. Such an environment can be provided initially in an ST device with the JET-level plasma conditions (Q=0.86 in Hot-Ion H-Mode) providing 0.25 MW/m**2 in outboard fusion neutron wall loading, and subsequently at twice the JET conditions (Q=1.7) to provide 1 MW/m**2. Conservative high-q and moderate-beta plasma conditions are calculated for the FNSF to minimize plasma-induced disruptions and allow the delivery of the required neutron fluence of 1 MW-yr/m**2 and duty factor of 10%. Fully modular designs for all the chamber components, including the single-turn toroidal field coil center-post, allow component installation and replacement via remote handling, which is required for the research operations of FNSF. Since the device support structures are hidden behind the chamber components, the FNSF provides a ready upgrade path to the Component Test Facility (CTF), which will require more stringent fusion nuclear and operational capabilities. Details of the physics, engineering, and research prerequisites assessments for the FNSF will be reported.