Introduction
The basic task of magnetic fusion research - i.e. creating and confining
sufficiently hot and dense plasmas for a reasonably long time - was to
a large degree resolved in the 20th century. In particular, the "scientific
feasibility of fusion" was demonstrated at JET and TFTR tokamaks
in their experiments with deuterium and tritium fusion fuels (further
details).
In the early 21st century, the next step tokamak ITER (Fig. 1) and
the accompanying research projects will now have to prove the technological feasibility
of fusion as a potential energy source.
With this mission objective,
fusion research is literally entering a new era in which the key role
will be played by technology research for future fusion reactors. Materials
need to be selected, capable of withstanding extreme thermal and mechanical
stresses in intense neutron radiation fields. Moreover, it is desirable
that materials used in fusion reactors should have as low as possible
activation from irradiation by fusion neutrons, and that any such induced
activity decays in a reasonable time scale. Tritium breeding from lithium
and the full fuel cycle have to be demonstrated and optimised. Plasma
heating sources as well as superconductive coils need further development.
Undoubtedly fusion technology research will be no less complicated
than the previous research into magnetic confinement. However the fact
that technology research is now required gives a clear indication of
the progress achieved in fusion and on the actual scale of available
fusion power. With the unique role that JET has been playing in this
progress it is an ideal place to pursue some of the necessary technology
tasks. One important aspect of fusion technology that JET contributes
to is in the realm of Remote
Handling. Other examples
follow below.
Increased use
of the JET facilities for Fusion Technology Research and Development
in preparation for ITER was one of the key objectives assigned to EFDA in
1999. For this purpose, a dedicated Task Force on Fusion Technology
was set up at JET in 2000, which has a close working relationship with
the broader EFDA
Technology Programme. Over the last five years, this
Task Force has launched a large variety of activities involving several
European laboratories. By presenting a few detailed examples of the
topics under research at JET we hope to demonstrate that in hi-tech
experiments, progress is achieved by careful, patient work rather than
in big strides. At the cutting edge of the current technology, where
new materials are tested under extraordinary conditions, improved performance
has to be acquired gradually.
Plasma-facing components and tritium introduction
JET provides invaluable expertise for the whole fusion community due
to its unique capability to operate with tritium, the heavy radioactive
hydrogen isotope. To find how and where tritium can be trapped inside
JET and to determine the characteristics of erosion and deposition
of the plasma facing components, investigations are carried out based
on the analysis of tiles or flakes removed during shutdowns or on direct
co-deposition monitoring (using e.g. quartz microbalances or rotating
collectors). The results of these activities are also used in the modelling
of the impurity transport inside the JET torus.
In many current tokamaks
- including JET - Carbon Fibre Composite (CFC) tiles act as the plasma
facing material. The fusion fuel, i.e. hydrogen isotopes, are co-deposited
together with carbon, beryllium and other elements present in-vessel
on these tiles. The co-deposits can fragment off to form flakes, which
in JET fall into sub-divertor zones close to the water cooled louvres
adjacent to the inner divertor (left end of tile 4 in figure below).
Flakes are collected via a remotely operated cyclone vacuum cleaner
and analysed. They have an average diameter of 0.4 mm and are saturated
with hydrogen isotopes. Optical spectroscopy reveals a layer structure
coming from a sequential deposition process.
Investigations of plasma exposed surfaces
Plasma exposed surfaces are investigated to provide data for understanding
and modelling the impurity transport in the plasma edge region (Scrape-Off
Layer, SOL), and the material erosion and deposition processes inside
the vessel.
The interaction of plasma with the CFC plasma facing tiles (Fig. 2) is the
major source of free carbon in the plasma, while Beryllium Evaporators,
used periodically mainly to reduce the amount of oxygen impurities
in the plasma and improve plasma conditions, represent the primary
source of beryllium in
JET. Carbon and beryllium are transported towards the upper tiles of
the inner divertor (tiles 1 and 3 in the figure) where beryllium is
stacked and carbon, after deposition, is re-eroded through chemical
sputtering and transported towards the inner flat tiles (tile 4).
The divertor tiles exposed in JET in the 1998-2001 campaigns have
been used to assess the amount of beryllium and carbon deposited at
the plasma facing materials. Secondary
Ion Mass Spectroscopy (SIMS) depth profiling has been made from
a number of samples on inner divertor tiles 1, 3 and 4.
The deposit
forms two layers on tiles 1 and 3. The outer layers (~2-6 µm
thick on tile 1 and 10-16 µm on tile
3) contain mostly carbon together with deuterium and a smaller amount
of beryllium. The films underneath the surface layer are very rich
in beryllium (~2-14 µm on tile 1 & 12-21 µm on tile
3). The measurements allowed the estimation of the amount of beryllium
on the tiles 1 and 3 and thus the calculation of the total amount of
beryllium deposited at the inner divertor: 22 ± 9 g. Unlike
tiles 1 and 3, very little beryllium was found in the ~85 μm thick
film on tile 4 in the shadowed region, where almost only carbon, with
very high deuterium content, and a well-marked interface to the carbon
fibre composite substrate has been observed. Similar investigations
have been carried out for the tiles of the outer divertor (6, 7 and
8) and, in general, the deposition patterns of fuel atoms, beryllium
and carbon showed much less heavy deposition and fuel accumulation in
the outer divertor than in the inner. This was not expected from classical
modelling of erosion/deposition. The asymmetry in the JET deposition
pattern could be explained by increased carbon erosion by the plasma
in the main chamber and sputtering at the inner divertor surfaces.
Cleaning of plasma-facing components
To avoid excessive tritium retention during future ITER operation, in
situ detritiation to be performed during operation-free periods would
be useful. Detritiation processes based on lasers or flash lamps are
being investigated at JET. After very promising results obtained on simulated
layers in the laboratory, showing a possible cleaning rate of more than
3m2/hour for a 50µm thick deposit, a flash lamp mounted on the
JET Remote Handling arm has been used for in vessel tests (see Figs. 4 & 5 and movie). The technical feasibility of this technique in a
tokamak environment has been demonstrated, and its efficiency is being
assessed by Ion
Beam Analysis, calorimetry and full combustion.
Laser cleaning of the plasma facing components via
layer ablation is also very promising. Ablation (erosion) of 50µm
of deposited layer on CFC (Carbon Fibre Composite) was obtained in laboratory
studies using a high frequency laser (output power 20W, 2J/cm2,
wavelength 1052 nm). A surface of 10x10 mm was automatically ablated
at 0.2 m2/hour,
without damaging the graphite substrate (Fig. 6). Extrapolation of these results
predicts that a 100W laser would have a removal efficiency of
1m2/h
for a 50µm
co-deposited layer in air. Work is ongoing to develop and test a laser
facility suitable for JET's Remote
Handling.
Management of tritiated materials
In fusion devices operating with tritium, different
tritiated materials are produced. Two main strategies can be adopted
for tritiated waste management: waiting for natural decay of the radio-nuclides
or applying some detritiation process. The second strategy is being investigated
by the Fusion Technology Task Force. Dedicated procedures for decreasing
the tritium content inside the materials removed from the torus are being
developed for stainless steel, carbon-based materials (graphite and carbon
fibre composite), organic liquids (pump oils, liquid scintillation cocktails)
and water, together with process and housekeeping wastes. In all these
projects, the right balance between the production of secondary waste
and the reduction of waste classification (see page 10 of the IAEA
safety guide), has to be reached.
Oxidation has been used to transfer the tritium atoms from organic oil
molecules to more stable and more easily treatable inorganic molecules.
Thermal desorption in the range from 20 to 1100°C under a stream
of helium containing 0.1% hydrogen has been used for carbon samples obtained
from tritiated JET tiles.
Heating of full CFC divertor tiles via radio frequency has been performed.
The amount of tritium before and after the procedure is being measured
by calorimetry and full combustion. Autoradiography (a method of detecting
and measuring the deposition, distribution and quantity of a radioisotope
present on any material by registering its radiation on a photographic
plate placed directly on the material) showed that after several heating
cycles at the average temperature of only 490 °C, more than 99% of
the tritium can be efficiently removed from a the surface of a tile (see Fig. 3).
Full combustion measurements showed that 95% of tritium from the bulk was
released.
Desorption tests have been also performed in a furnace under
a stream of argon gas containing 5% of hydrogen. These experiments showed
that the optimal detritiation temperatures are between 300 and 800°C
and decontamination factors (i.e. initial activity / final activity)
between 20 and 90 can be obtained.
For stainless steel the studies have been performed with
the oxidation method on samples from the Belgian SCK-CEN
laboratories and a French fast breeder fission reactor. Large samples (250 to 700
g) were used in order to determine the impact of the treatments on tritium
trapped both at the surface and in the bulk. Using smear tests to evaluate
the residual surface tritium contamination, a decontamination factor
of about 210 was obtained. However, further developments with measurements
of tritium content in the bulk material are needed to fully determine
the efficiency of the process.
Other samples were treated in a furnace
up to 1100°C
in air, or in argon with 5% hydrogen. Heating the samples for 3 hours at
400°C led to a decontamination factor of about 5 in air and 8 in argon
with hydrogen. The factor increased respectively to about 130 and 110 when
heating at 1000°C for half an hour.
The system for water detritiation is based on tritium enrichment
in a Liquid Phase Catalytic Exchange column of the contaminated water from
the processing of the operational gases in JET's Active Gas Handling System.
This water is then dissociated in oxygen (discharged in the atmosphere) and
a mixture of hydrogen isotopes in an electrolyser. Hydrogen (H), Deuterium
(D) and Tritium (T) are then separated by Cryo-Distillation (method based on
the different volatility that decreases from H2, through HD, HT, D2, DT to
the molecule T2 ) and Gas
Chromatography.
The design of a fully integrated plant as well as the testing of all its key
components has been carried out as part of research and development in preparation
of the ITER plant and could be directly applied at JET.
Any active and/or toxic waste is either stocked on site, or safely disposed
of. Even though the UK law is very strict concerning any such waste, JET imposes
its own more demanding internal targets on safety. Both technical and scientific
staff are well aware of our responsibility to keep the environmental impact
of our research as low as possible.
Vacuum pumping and gas handling
The design of the ITER high vacuum system is based on a number of supercritical
helium cooled cryosorption
pumps providing a high pumping speed and capacity,
as well as fast on-line regeneration. To pump helium, which cannot be
condensed, and to help to pump hydrogen, the pumping cryopanels are coated
with activated
charcoal granules. Activated charcoal is a highly porous
carbon with millions of tiny pores between the atoms, creating surface
areas of several hundreds of square meters per every gram of charcoal,
so that it has a unique adsorption capacity. After preliminary tests
at FZK,
Germany, a large scale test arrangement was built at JET in the
Active Gas Handling System (Fig. 7) to assess in detail the carbon-tritium interaction
and to derive performance parameters essential for the design of the
ITER cryosorption pumps. This new pumping cryopanel (Fig. 8) was first operated
under the JET Trace Tritium Campaign in 2003, pumping gas from the JET
torus and neutral beam injectors. It was observed that the pumping cryopanel
worked according to the design specifications.
The JET vacuum pumps, including this new cryopanel system, pump all
gases from the torus and other systems (e.g. Neutral Beam Injectors)
into the Active Gas Handling System, where the different hydrogen species
(H, D and T) are sorted out using isotope separation techniques, and
deuterium and tritium are stored for future JET fuelling (See Figs. 9 and 10) .
A new purification
system called PERMCAT (Fig. 11) is also being installed in the Active Gas Handling
System to remove impurity gases such as helium He, carbon dioxide CO2,
water H2O, or methane CH4 from the collected gases. Figure 12
shows a schematic of the system:
Pumped gas flows into the PERMCAT where tritium is exchanged with protium
(i.e. the common light hydrogen isotope H) through a palladium/silver
membrane.
Diagnostics studies - optical fibres
Optical fibres offer
an attractive practical solution to transport light through the complicated
geometry surrounding the fusion reactor. However they can suffer from
serious radiation-induced optical absorption and radioluminescence. Special
fabrication and glass hardening techniques must be used to deploy suitable
radiation-resistant fibre in a tokamak machine like ITER that produces
neutron and gamma radiation during plasma operations.
As JET is
the closest machine to ITER, including radiation flux due to fusion reactions,
studies have been undertaken to demonstrate the feasibility of using
optical fibres in diagnostics systems during reactor operation, and in
particular the possibility of using large diameter fibres, i.e. with
a core diameter of 0.6 mm, acrylate coating and suitable hydrogen treatment
to enhance radiation tolerance.
Special hardware was installed in the JET Torus Hall in order to test
this fibre during Trace Tritium Experiments in 2003. As a result, a small
but detectable loss in optical transmission due to radiation during plasma
discharge was observed. The optical loss was measured to be 6% at maximum.
When the radiation decreases the fibre recovers its transmission
capabilities totally, suggesting that no permanent damage has taken
place. The reserve of hydrogen implanted during the pre-treatment is
probably sufficiently high to repair the damage.
Direct measurements carried out in luminescence mode revealed the presence
of radioluminescence during the plasma pulse. Consequently, an increase
of the optical transmission following the shape of the pulse is observed
throughout the pulse. However, no correlation was found between the radiation
conditions and the luminescence intensity. This probably results from
the non-uniformity of the irradiation conditions.
Other Fusion Technology topics at JET
Fusion Technology research and development at JET comprises five main
topics, with substantial emphasis on tritium-related tasks (see Fig. 13). In addition to the investigations discussed above, parts
of the JET Facilities are also used as test beds for studying prototypes
for ITER, such as bypass switches for power supplies, or carbon-based
tiles under high ion loads. Moreover, after more than 20 years of operation
and experience with the use of tritium, beryllium and remote handling
for maintenance, JET provides a unique source of information which
helps ITER's design and licensing. Data is collected on component failure
rates in various sub-systems (vacuum system, heating systems, power supply,
active gas handling system) and on occupational radiation exposure (depending
on worker categories and operation conditions). Despite the fact that
the ITER design calls for a machine that is significantly larger than
JET and different operational procedures are expected, the raw data and
the analysis results obtained from its study are relevant and offer
significant insights. |
