EFDA Technology Work Programme 2005
Activities of 2005-2006
Safety and Environment (TS)
Task area: Waste Management (TSW)
TW5-TSW-001-D2
Precondition of the
dose rate limits for
recycling
Principal Investigator:
Scientific Staff:
L. Ooms
S. Leblanc
S. Boden
D&D
SCK•CEN, Mol, Belgium
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276/06-06
October 2006
Distribution List
J. Pamela
M. Gasparotto
W. Gulden
V. Massaut
S. Ciattaglia
P. Sardain
Y. Poitevin
EFDA Leader
EFDA Associate leader for technology
EFDA-CSU Garching, Safety & Environment
EFDA-CSU Garching, Safety & Environment
EFDA-CSU Garching, Safety & Environment
EFDA-CSU Garching
EFDA -CSU Garching
L. Di Pace
R. Pampin
K. Broden
L. Sponton
R. Bestwick
O. Gastaldi
ENEA- Frascati
UKAEA- Culham
STUDSVIK
STUDSVIK
AMEC-NNC
CEA
M. Decréton
G. Collard
L. Noynaert
F. Druyts
I. Uytdenhouwen
J. Braet
K. Dylst
S. Boden
SCK•CEN, Fusion Coordinator (3 copies)
SCK•CEN
SCK•CEN
SCK•CEN
SCK•CEN
SCK•CEN
SCK•CEN
SCK•CEN
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L. Ooms
S. Boden
Author
For verification
M. Decréton
Fusion Coordinator
For approval
2
TABLE OF CONTENT
1.
Introduction ......................................................................................................... 4
2.
Reference Model AB ........................................................................................... 5
2.1. Plant Model AB general description .......................................................... 5
2.2. Plant Model AB main parameters .............................................................. 7
2.3. Different IVC's and materials used............................................................. 8
2.4. Activation analysis ..................................................................................... 9
2.5. Time Scale ................................................................................................ 10
2.6. Model B .................................................................................................... 11
3.
Actual activity limits (experience in fission) ..................................................... 12
3.1. Recycling (melting) .................................................................................. 12
3.2. Hot cell facilities ....................................................................................... 14
3.3. Waste issue during dismantling ................................................................ 16
4.
Recycling process of fusion material ................................................................ 18
4.1. Fusion classification ................................................................................. 18
4.2. Recycling possibilities of the materials .................................................... 19
4.3. Remote handling feasibility ...................................................................... 21
4.4. Tritium handling during recycling............................................................ 22
5.
Conclusions ....................................................................................................... 24
6.
Acknowledgement ............................................................................................. 27
References .................................................................................................................. 28
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1. INTRODUCTION
The radioactive waste treatment of future fusion reactors becomes more and more a
strategic issue. Although fusion reactors have the advantage of producing materials
with low radio-toxicity, but in large quantities, it is important to demonstrate that
recycling is an option for a maximum reduction of fusion waste.
Previous reports have described in detail the techniques applied today for recycling
materials out of the fission industry. This report will precondition the limits for
recycling of fusion materials, mostly based on dose rate, but other aspects being also
taken into account. The basis for this preconditioning is the experience in fission
technology (eg. dismantling, recycling, remote handling, regulations...) and the
expected properties (eg. dose rate, heat production) of the different In Vessel
Components (IVC) to be recycled.
These limits must allow us in a next stage to:

quantify the cooling period before recycling

define what to do with the waste coming out of the recycling process

define the consequences for the recycling strategy of In Vessel Components.
In this deliverable the limits for recycling of fusion materials are emphasized in all
aspects of the recycling process (dismantling, melting, fabrication, waste treatment,
intermediate storage, etc). The study will concentrate on remotely operated
recycling, as the preceding reports stated that this is unavoidable when highly
active plasma facing components are considered. The recycling strategy of a fusion
plant can be further studied, based on the activity limits proposed in the document.
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2. REFERENCE MODEL AB
As a reference for this document the properties and parameters of the Power Plant
Conceptual Study (PPCS) plant model AB are used. In this chapter an overview is
given.
2.1. Plant Model AB general description [10]
Model AB is based on Helium-Cooled Lithium-Lead (HCLL) blanket which uses
EUROFER as structural material, Pb-17Li (Li at 90% in 6Li) as breeder, neutron
multiplier and tritium carrier, and helium as coolant.
A helium cooled divertor featuring a maximum heat flux of 10 MW/m2 as in
Model B is considered as reference solution and a First Wall (FW) heat load limit of
0,5 MW/m² has been assumed.
Main design rational and assumptions
For the reactor Model AB, the following basic guidelines have been considered:

the breeding blanket main parameters and architecture are based on the
corresponding HCLL blanket;

the shield is divided into two regions: high temperature shield (HTS),
also called manifold zone, and low temperature shield (LTS).

He and Pb-Li collectors are integrated in the HTS, in order to cool HTS
through the He collector.
Design of the in-vessel components
Blanket and shields segmentation: The hypothesis of the large sector handling has
been assumed, in order to reduce the number of segments to be removed and the
number of mechanical and hydraulic connections to be disconnected and reconnected
during the blanket maintenance phase and then increase the availability of the
machine.
The blanket modules have dimensions of about 4 m (poloidal) x 2 m (toroidal)
20 modules (comprising the one in the port) cover a 40°sector, leading to
180 modules in the entire reactor.
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The vertical sections of the reactor describing the blanket modules segmentation is
shown in figure 1.
Figure 1: Model AB - Vertical section (in correspondence of the equatorial port) [10]
Blanket modules: the HCLL blanket is based on the use of EUROFER as structural
material, of Pb-17Li (Li at 90% in 6Li) as breeder, neutron multiplier and tritium
carrier, and of He as coolant.
A generic HCLL blanket module of 2 m (poloidal) x 2 m (toroidal) x 1m (radial) is
depicted in figure 2.
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Figure 2: Model AB – Generic HCLL blanket module [10]
BP: Back Plate
FW: First Wall
BU: Breeder unit
The Pb-17Li breeder slowly flows throughout the box for allowing external tritium
extraction.
Shields and Vacuum Vessel: The shield is split into two parts, the HTS and the
LTS. The thickness of the HTS is 30 cm in the inboard and 35 cm in the outboard,
and the thickness of the LTS is 20 cm in the inboard and 35 cm in the outboard.
Both the He and Pb-17Li manifolds are integrated in the HTS made of EUROFER.
To improve shielding efficiency the LTS is made of EUROFER and Tungsten
Carbide (WC) and is cooled by water.
2.2. Plant Model AB main parameters
The PPCS model AB is a "near term" fusion power reactor based on limited
extrapolations of present-day plasma physics and technology knowledge. Therefore
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the reactor parameters of the plant model AB are estimated using the same plasma
physics assumptions as for the models A and B.
Table 1 compares the most important parameters of the reactor models A, B and AB
in order to pinpoint the differences. Anyway, if a parameter of the plant model AB is
not available, the one of the plant model B will be used as a best estimate (fi. Coil
dimensions).
Parameters
Model A
Model B Model AB
1,55
1,33
1,5
5
3,6
4,24
Major Radius (m)
9,55
8,6
9,56
Average neutron wall load
(MW/m2)
2,2
2
1,84
Divertor Peak load (MW/m2)
15
10
10
Unit Size (GWe)
Fusion Power (GW)
Table 1: Parameters of the plant models A, B and AB
2.3. Different IVC's and materials used
Cooling
Structure
Breeding Function
The materials used in each PPCS Model are shown in table 2.
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PPCS Models
Materials
Model A
WCLL
LiPb
Blanket
Model B
HCPB
Model AB
HCLL
Blanket
Blanket
Blanket (n multiplier)
Blanket
HTS
Blanket
Divertor
LTS
EUROFER
In vessel shield
In vessel shield
Blanket (+ FW)
VV
VV
VV
SS 316
HTS
ZrH
Divertor
CuCrZr
Divertor
Divertor
Divertor (armour)
W
LTS
WC
Divertor
W alloy
Blanket
Divertor
LTS (cooling)
Water
VV
Blanket
Blanket
Divertor
Divertor
He
VV
HTS
Table 2: Materials used in the PPCS Models
Li4SiO4
Be
8
The materials have been classified in three main groups according to their function:
breeding, structure and cooling. Table 2 indicates in which component the material
features, model by model. Other or additional functions of the materials are
mentioned between brackets.
2.4. Activation analysis [1]
The report TW4-TRP-002 D2e describes the activation calculations for the PPCS
model AB in detail. The results, used in this report, are summarized in the table
below. Indeed the dose rate of the materials to be recycled is one of the most
important characteristic of the fusion materials taken into account for remote
handling.
Component
FW armour
Blanket FW
Blanket
Breeder
Blanket
Manifold
HTS
LTS
VV
TF coil
Divertor
T0
Back
Front
2.67E+04
8.77E+04
2.72E+02
3.67E+04
T0 + 10 years
Back
Front
4.19E+00
1.70E+01
1.47E-01
7.93E+02
5.85E+02
3.22E+02
1.41E-02
3.17E-08
1.78E+03
9.41E-01
4.00E+00
3.10E+00
3.07E-01
7.23E-04
1.30E-09
T0 + 50 years
Back
Front
4.67E-02
3.81E-02
2.74E-03
1.60E-02
2.09E-02
1.62E-02
2.93E-03
4.05E-06
3.22E-11
1.74E+04 5.60E+00 3.16E+00 2.92E-02
2.82E-02
Table 3: Dose rate of fusion materials (Sv/h)
T0 + 100 years
Back
Front
8.68E-03
1.04E-03
2.21E-04
3.92E-03
1.63E-04
1.01E-04
1.78E-04
1.48E-07
2.56E-11
1.59E-04
3.34E-03
The dose rates given in the table above take into account the exposure time of each
component. The irradiation time of the components are shown in table 4.
Component
Divertor
Blanket (incl. FW and armour)
High Temperature Shield
Low Temperature Shield
Vacuum Vessel
TF Coil
Replacement schedule
Each 2,5 years
Each 5 years
Each 5 years
Life time component (± 29 years)
Life time component (± 29 years)
Life time component (± 29 years)
Table 4: Exposure time of the different components
A point to mention here is the important role of impurities in the different materials,
which can shift a material from one category to a higher one due to the activation
products of these impurities [16].
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2.5. Time scale
In this paragraph we summarize the ideas expressed in different reports concerning
the strategy in different steps of the material handling of fusion plants. The different
steps can be summarized as follows:

Storage after irradiation

Recycling (all steps from separation of components to refabrication)

Storage awaiting reuse

Reuse.
The storage [2] after the components are removed out of the fusion reactor has
several goals:

Removal of the decay heat.

Desorbtion of tritium: a part of the remaining T in the In Vessel
Components will be trapped in the water if wet storage is applied even if a
post-shutdown degassing procedure is performed.

Removal of Tokamak dust: the dust is collected prior to the removal of the
IVC's, but some dust (order of magnitude = a few kg) remains on the
surface and will be released in the storage facility and has to be trapped.

Attenuation of gamma irradiation: the storage aims to reduce the dose rate
with a factor 10.
Whether this storage is wet or dry has no direct importance in this document, it
will only influence the operating conditions of the storage facility. More
important is the cooling time to determine the initial dose rate of the
components for recycling.
The reference [2] describes also the different stages in the storage of IVC's:

Stage 1: Temporary storage over a time period of maximum 5 years. The
storage will be performed in the fusion plant and its major goal is to reduce
the dose rate with about 1 order of magnitude.

Stage 2: Interim storage over a period of 100 years. The major goal for this
on site storage is to reduce the dose rate (a few orders of magnitude) in
order to perform off-site transportation.

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Stage 3: Final disposal (off site).
10
In this document we study how to avoid stage 3 and decrease the time period of
stage 2 of the strategy mentioned above by applying recycling of the materials.
After the first storage period (at least 5 years) the recycling process (from seperation
to refabrication) can be started. The method to be applied and the necessary shielding
depend on the materials used to fabricate the components and their dose rate. The
different material paths will be studied in the next report.
After fabrication and before reusing the components, a second storage period
(awaiting reuse) will be necessary in order to perform a full change of components
during a shut down of a reactor.
2.6. Model B
The model AB was the reference for the study, although the breeding blanket of the
model B would also be regarded for the TSW001 task. For this study the breeding
blanket (HCPB) containing Beryllium in the breeder zone has no influence on the
conclusions for the preconditioning of the limits. The presence of Be in the blanket
only confirms the necessity of confinement measures to recycle fusion components
due to the toxicity of Be.
Nevertheless we can mention here the experience in Be-recycling at SCK•CEN [17].
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3. ACTUAL ACTIVITY LIMITS (EXPERIENCES IN FISSION)
3.1. Recycling (melting)
In the previous report [3] melting facilities responsibles have declared that
materials with an activity less than 1000 Bq/g can be hands on recycled in
current melting facilities. Other documents refer to the dose rate of 10 µSv/h as
limit for the hands on recycling. The goal in this paragraph is to compare these
values and to perform some simulations in order to assess the limits for
recycling.

1000 Bq/g
The 1000 Bq/g can be converted to a dose rate using a very simple
method (1 kBq 60Co = 4 µSv/h at 1 cm) [6] [7].

10 µSv/h
The same exercise can be done for this value. If the dose rate of
10 µSv/h is received by an operator (full exposure = 1600 h to 2000h
per year), his annual dose will be 16 mSv to 20 mSv, which is again
below or equal the actual limit for nuclear workers (20 mSv/y following
ICRP 60 recommendation).

Simulation 10% exposure (or 90% shielding)
The experience in decommissioning learns us that full exposure is never
reached. On the contrary the theoretical dose estimate is very
conservative, therefore only 10% of the full exposure time is taken into
account (or 10% of the dose during full exposure).
If we calculate the dose rate of the material with these assumptions
(10% exposure and a limit of 20 mSv/y), the dose rate of the material
could increase up to 125 µSv/h.
This rudimentary simulation is not suitable to define new limits for recycling, but
some conclusions can be drawn:
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
The limits used or declared are conservative, and they are currently
applied in fission recycling (existing melting facilities). Anyway they
meet the only requirement which is the annual dose rate of the operators
of the facilities.

A simple simulation taking into account a reduced exposure (or
shielding) shows that the dose rate of the materials could rise with at
least a factor 10.

The gain can not only be influenced by decreasing the exposure time,
also shielding or simply the distance can reduce the dose uptake of the
operators.

Authorities have accepted the limit of 10 µSv/h and have the tendency
to reduce limits instead of raising them.
Taking into account all these statements, the 10 µSv/h should be the limit for
hands on recycling. But classifying the materials that have a higher dose rate
directly into remote recycling is not correct. Therefore a new category should be
added which is treated in such a way that the annual dose rate of the workers is
respected by applying reduced exposure time, shielding and other measures.
We propose to define this category as "shielded recycling". This is further
discussed in paragraph 4.1.
Remark: Relation between dose rate (contact or 1m) and activity in Bq/g
The relation between dose rate and activity in Bq/g might be obvious for simple
cases where only 1 gamma ray emitting radionuclide (e.g. 60Co) is present. In this
case the relationship is mainly influenced by the geometry of the structure that is
being measured. Activity distribution and self shielding are in this case
important.
In reality, there are three important elements to take into account in case of
measuring a dose rate and trying to calculate the activity in Bq/g (or vice versa):

A specific dose rate value can lead to different specific activity (Bq/g)
depending on the composition of gamma ray emitting radionuclides (e.g.
100% 60Co versus 50% 152Eu+154Eu and 50% 60Co).
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
Pure beta emitters with long half lives (e.g.
55
Fe,
59
Ni,
63
Ni) that might be
present in high concentrations do not cause an increase in the external dose
rate but well to internal contamination.

The big difference in half lives of the different radionuclides leads to a
changing relation with time.
This results in the following:

Using dose rates to draw the line between Hands On, Shielded and Remote
Handling seems more logic than using specific activities.

The environment where recycling is applied must have the necessary
confinement to avoid internal contamination of the workers.

However, one should use specific activity (Bq/g) values and the index
concept for all materials that might be unconditionally or conditionally
released. It has to be noted that slightly activated materials that have a dose
rate < 1 µSv/h after a certain cooling time might still contain concentrations
of pure beta emitters exceeding the release limits (therefore the 1 µSv/h level
does not correspond to any change in category).
3.2. Hot Cell facilities
In the nuclear world a lot of hot cell facilities are in operation, each with their
own specialisation. A few examples have been described in this paragraph,
showing that remote handling in a hot cell facility is a well known practice.
To meet the requirements for radiation protection of the fusion materials to be
recycled, manipulation in a hot cell is necessary for a part of the materials.
Since the dose rate is only a parameter for the design of a hot cell facility, the
thickness of the shielding can be adapted to the dose rate of the materials to be
handled.
At the nuclear sites at Mol and Dessel several hot cell and shielded facilities are
available, all of them very specialised for their purposes. But looking at the
activities and dose rates which can be handled in these facilities, one can
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conclude that fusion material can be handled in these facilities for what
concerns this topic.
Hot cell facilities at SCK•CEN (specialised in  contamination):
•
Laboratory for Low and High Activities (figure 3)
•
Hot cell BR2
Building for Vitrified waste at Belgoprocess
•
Dose rates up to 1.4 E4 Sv/h.
•
Treatment of canisters
Figure 3: Laboratory for Low and High Activities
When comparing the value with those in table 3, we see that at T0 we have a
maximum dose rate of ± 9 E4 Sv/h, but the dose rate of the materials decreases
rapidly (days) to values well below the limits of the vitrified waste building of
Belgoprocess.
Taking into account a temporary storage (see point 2.5) of at least 5 years, fusion
materials could be handled in installations with comparable measures as used in
existing hot cells.
Remark: the dose rate of the fusion materials are calculated for the cycle of each
component. When recycled material is reused, an important build up of activity
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will be generated which has to be taken into account in the design of the
shielding parameters.
Taking a closer look to the services available in a hot cell facility, one can state
that these facilities are very specialised in what they are doing, so not directly
applicable for fusion materials, but state of the art techniques are applied with
remote handling.
The following remote operations are performed in existing hot cell facilities:
•
Milling
•
TIG Welding
•
Liquid penetrant inspection
•
Video inspection
•
Decontamination
•
Hydrostatic Pressure Testing
•
Dimensional verifications (Laser profilometers, contact micrometers, dialgauge micrometers…)
•
Destructive mechanical testing (tensile, charpy edge, fatigue tests…)
•
Non-destructive testing (e.g. Ultrasonic testing…)
•
Camera guided cutting operations, cold and hot techniques (e.g. sawing,
EDM…)
•
Microscopic examinations
•
Metallographic sample preparation
•
Atmospheric conditions (temperature, pressure, environment (inert gas,
water…)
•
High temperature furnaces, vacuum heat treatment furnace
•
Laser profilometers, contact micrometers, dial-gauge micrometers.
Precisions up to ± 10 µm are noted for remote machining of precise test
specimens of small scale (a few cm). Of course the accuracy depends on the
installation and the manipulators installed (classified in levels).
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3.3. Waste issue during dismantling
Even if all In Vessel Components of a fusion plant can be remote handled, a
certain amount of waste will be left overcoming out of the recycling process.
This amount of waste must be disposed of. In this paragraph some experience
of waste generation in function of the applied cutting technique is given. With
this information we try to estimate in the next studies a percentage of waste
generated during recycling.
During the dismantling of the BR3 reactor a lot of experience is gained on
cutting techniques in all aspects (e.g. secondary waste production, cutting
speed, easiness of handling, remote operation feasibility…).
Also special tests were performed to compare the secondary waste production
of different cutting techniques. Table 5 gives a comparison between a
mechanical cutting technique and a plasma cutting technique.
Test number
Mechanical cutting
Mechanical cutting
Mechanical cutting
Kerf
width
(mm)
1.2
1.2
1.2
Kerf
Surface
[cm²]
890
847
661
Swarfs or slags
volume
[cm³]
107
102
79
Plasma cutting
3
199
60
Plasma cutting
3
359
108
Plasma cutting
3
347
104
Table 5: Waste production of different cutting techniques
Ratio
Volume /
surface
0.12
0.12
0.12
0.30
0.30
0.30
It is difficult to define a percentage of waste mass in function of the mass of the
materials to be treated, since the number of cuts necessary for dismantling and
their surface will define the amount of secondary waste generated during this
operation. But the comparison of the two techniques let us conclude that
plasma cutting generates twice the amount of waste than mechanical cutting.
Cold tests on the reactor internals in the past [8] confirm these values, that hot
techniques (EDM and plasma cutting) generate more secondary waste (up to 5
times more) than cold (mechanical) techniques.
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4. RECYCLING PROCESS OF FUSION MATERIAL
4.1 Fusion classification
Reference [4] compares the different limits of the different classification methods
applied on fusion material. Figure 4 gives an overview.
Figure 4: Waste classification methods: comparison [4]
The upper limit for the category of SRM was set to 2 mSv/h, the origin of this
value can probably be found in the IAEA documents (Safety series No. 111-G1.1: Classification of radioactive waste, reference [5]) and is defined as (point
318): the 2 mSv/h distinguishes the Low Level waste class and the intermediate
level waste class, saying that it is waste with low heat dissipation but require
shielding during normal handling and transportation.
The statement is based on following references:

IAEA doc Safety Series No 54 (1981): Underground disposal of radioactive
Wastes: Basic Guidance

IAEA doc Safety Series No 6 (1985): Regulations for the Safe Transport of
Radioactive Material.
In the first document only the classification is described (no numbers), the second
describes the transport norms where the 2 mSv/h is the limit for the contact dose
rate outside a transport container.
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Taking into account all statements noted above, we propose to adapt the diagram
in figure 4 to the one given in figure 5. Two changes have been applied:
1.
The category shielded handling is added. This category divides the
SRM category into two (hands on recycling and shielded recycling).
2. Materials with a dose rate and/or a heat production such as the limit
values of the categories CRM and PDW do not pose problems for
handling in a hot cell facility. Even existing facilities can handle
materials with those kinds of dose rates.
3. Attention should be paid to heat removal if high values (e.g. > 1000
W/m3) are encountered
1 µSv/h
2 mSv/h
20 mSv/h
< 10 MW/m³
< 10 MW/m³
< 10 MW/m³
NAW
SRM
CRM
> 10 MW/m³
(1)
PDW
10 µSv/h
Hands on
Shielded
Recycling
(2)
Remote Handling
(1) PPCS old classification
(2) New proposal
Figure 5: Waste classification methods: new proposal
4.2. Recycling possibilities of the materials
As noted in previous studies, materials out of the nuclear industry which are
recycled (melted) at present have a low specific activity (< 200 Bq/g). The
recycling process is performed after the materials are cleared or below a
conditional clearance limit. Conditional clearance obliges the reuse of the
materials in the nuclear industry. To achieve conditional clearance often a decay
period of a few years (up to 10 years) is applied to reach the limits.
Up to now the materials were reused in the applications listed below:

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Shielding plates (up to 90% of recycled material)
19

Transport containers for shipping and storage of low level and high level
radioactive waste (MOSAIK® container: up to 25% of recycled material).

Monolith cast iron container for final disposal of high level radioactive
metallic waste (e.g. core internals, reactor pressure vessel segments…).

Granulates made of recycled material are mixed with concrete to fabricate
shielding containers, or a variant with an inner and outer lining filled with
a mix of granulates and concrete.
If we put this experience in perspective to fusion materials where a dose rate up
to a few mSv/h is to be expected (after a decay period of 100 years), reuse of
materials is not obvious due to the high dose rate and other technical aspects have
not been taken into account yet!
Since all the materials fabricated with melted materials are used for shielding or
enclosure of nuclear materials, a high dose rate of these materials is not allowed.
Therefore reusing materials coming out of fusion reactors who have a high dose
rate should be submitted to longer (>100y) decay storage or refabricated to reuse
in fusion power plants as IVC. The last possibility implies that the fabrication of
this parts have to be fully remote controlled as mentioned in previous reports, the
feasibility of this must be analysed in detail.
Another aspect which is not yet discussed at the moment is the secondary waste
generated during the melting process. The experience at BR3 shows that one
must count on a waste fraction of 2 to 3% of the total mass to be melted. This
fraction will most probably change if we apply the melting process on other
materials, with a much higher dose rate and other characteristics. Melting
facilities indicate a waste production of <5%, this value will be used to define the
material cycle in the deliverable D7.
The waste coming out of the melting process consists out of dust (on filters),
slags, debris, refractory components, etc… The melting process is also regarded
at as a decontamination of the base material, but decontaminating the base
material results in a concentration of radioactive material in the waste. Therefore
these materials must be treated as radioactive waste as in fission industry. This
has also consequences for the waste produced if fusion materials would be
melted. Indeed the waste fraction will have a much higher dose rate, so they will
be classified in a higher waste category.
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Remark: Melting in fission and fusion area: difference in end results
Radioactive waste originating from fission reactors, especially metal
components from various circuits, is in many cases a mixture of typical
fission products (e.g.
137
Cs,
90
Sr), potentially actinides (U, Pu,…) both
coming from contamination and activation products (e.g. 60Co, 55Fe, 59Ni,
63
Ni). After melting many of the fission products and actinides are
concentrated in the secondary waste fraction. Many of the typical
activation products (with atomic number close to Fe) alloy in the end
product. Therefore, the net result is an end product containing many of
the activation products in a more homogeneous distribution than before
and a secondary waste fraction containing many of the concentrated
fission products. If the concentration of activation products is low enough
the end product can be released and reused. We can conclude that melting
of fission waste is very efficient and effective if the original waste is
containing a high amount of fission product and very low amount of
activation products.
In the case of melting fusion elements, the end product would probably
still contain the majority of the activity as compared to the original
element and we would probably not gain a big reduction of the dose rate.
4.3. Remote handling feasibility
As discussed in previous paragraphs the dose rate of the In Vessel Components is
not a problem for existing hot cell facilities. Nevertheless a closer look has to be
taken on the methods and techniques applied during recycling, starting from
dismantling to fabrication. A lot of techniques have been applied already in a hot
cell environment, with a very good precision (± 10 µm) for very small pieces as
described in paragraph 3.2. Also testing the fabricated parts is a well known
technique under remote operation. Even melting is applied in a hot cell
environment (eg. Vitrification) but requesting very complex installations.
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Finally we must keep in mind the danger of inhalation or ingestion for the
worker, therefore the environment were recycling is applied must have the
necessary confinement to avoid internal contamination of the workers
4.4. Tritium handling during recycling
Even if detritiation is performed in the fusion plant (mainly surface detritiation),
a certain amount of tritium will be captured in the metal matrix. Therefore special
attention must be paid to the tritium behaviour during the different recycling
processes.
If the tritium containing material will be melted during its recycling, the metal
will release a certain amount of tritium, since melting is the process for
detritiation of metals containing tritium [18] & [19].
The quantity of tritium release is determined by several factors:

the initial concentration of tritium in the metal

the possibility to use a carrier gas

the carrier gas used and its composition.
Some existing melting facilities (fission materials) have set up a limit for the
tritium concentration (e.g. Siempelkamp: lower than 2000 Bq/g is necessary
before melting).
If new melting facilities will be equipped with a tritium capture system in the offgass circuit, it should be possible to handle metals with a much higher
concentration.
A simplified schematic view of a tritium removal process in the off-gas circuit is
set up below. The different steps are:
1. Melting of the material and injection of a carrier gas (e.g. He)
2. Transport of the carrier gas and tritium over a column with CuO curls at
600°C. The H2, HT and T2 is converted to H2O, HTO and T2O. In practice
HTO and H2O (tritiated water).
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3. Distillation of the tritiated water out of the carrier gas.
HT
&
carrier gas
1
Carrier gas
&
HT
Metal
melt
Carrier
gas (He,
Ar, …)
2
3
CuO
curls
600 °C
HTO
&
carrier gas
carrier
gas
HTO
Figure 6: T-capture in the off-gas circuit
Of course this scheme is simplified, the interaction of other products in the offgas with the tritium removal system should be studied in detail.
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5. CONCLUSIONS
In this document several topics were discussed in order to make progress in the
feasibility study of the recycling of fusion materials. Indeed the dose rate of the In
Vessel Components (IVC) is after its exposure time very high, but decreases rather
quickly in comparison with fission material.
Nevertheless the feasibility of handling fusion components with a high dose rate is
studied in this document. The findings can be summarized as follows:

Existing hot cell facilities can handle materials with a dose rate equal to the
calculated dose rate of fusion material after a temporary storage of 5 years.

If fusion materials are reused, we must verify the activity build up during the
second and following irradiation cycles. Therefore we recommend studying
this phenomenon in order to have a guide number for the dose rate after
reuse. With these data the design of the hot cell facility can be adapted.

Existing hot cell facilities can provide a wide range of services in various
domains. A large number of techniques for fabrication and testing are
available and a precision of 10 µm can be reached at the moment for remote
machining of small pieces. Of course the fabrication techniques for the
fusion materials must be checked with the available technology and remote
control, if not available, new R&D is necessary to adapt its application in a
hot cell environment. Also the characteristics of the components (e.g.
dimensions, weight…) will affect the design of a facility. A co-operation
with manufacturers of the IVC's is necessary in the future to discuss this
topic.

The internal contamination hazards (inhalation, ingestion) defined by the
specific activity implies confinement restrictions for all the facilities. Indeed
pure beta emitters or very low energy gamma are also of importance; and
these nuclides can have rather long half lives (like e.g.
14
C, 92Nb, 55Fe, 63Ni,
etc or even tritium). These radionuclides play a role in todays waste
classification, and should be taken into account in the fusion approach.
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Melting is the most favorable process for detritiation of metals who contain tritium
in their matrix. Therefore a tritium capture system should be installed in the off-gas
system of the melting facility. A simplified scheme to perform such operation is
defined in this document. As indicated in other documents, melting is not the only
recycling solution (e.g. refractory materials and ceramics), but the same process of
T-capture can be installed in another environment, to separate the tritium out of an
air flow (f.i. ventilation system).
The most important task in this study was the definition of the activity limits for
handling. In this document we propose a slightly changed diagram, based on the
path or environment where the materials must be handled. This proposal can be the
start of the discussion regarding this topic. Three points have changed:

The former SRM category (simple recycled waste) is subdivided into two
groups:
o Hands on recycling: materials with a dose rate up to 10 µSv/h.
o Shielded Recycling: materials with a dose rate from 10 µSv/h to
2 mSv/h.

The materials with a dose rate above 2 mSv/h can be handled in a hot cell
environment, therefore the group "remote handling" is introduced
containing 2 previous categories namely:
o CRM: 2mSv/h to 20 mSv/h
o PDW: above 20 mSv/h

As clearance level is defined at international level [IAEA RS.G1.7 to be put
in the ref list], the dose rate of 1 µSv/h does not represent a border between
former NAW and SRM.
The role of the temporary storage is an important factor in the whole classification
of the materials since a longer decay storage will simplify the recycling process.
Possibly difficult recycling/fabrication techniques can be applied in a shielded
environment instead of remote controlled if a longer (few years or decades) decay
period is applied. This remote controlled or shielded environment must in anyway
fulfill also the confinement requirements.
This fits all in the strategy of recycling which should be defined based on current
information. In fission industry recycling is applied for low dose rate materials and
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the product is very basic (no difficult fabrication and testing techniques). Fusion
materials are the opposite; high dose rate, difficult fabrication and testing
techniques. Therefore further investigation should be performed on the fabrication
and testing methods necessary for fusion components and if these services can be
applied in a shielded or remote controlled environment, with the necessary
confinement measures.
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6. ACKNOWLEDGEMENT
This report, supported by the European Communities, was carried out within the
framework of the European Fusion Development Agreement. The views and opinions
expressed herein do not necessarily reflect those of the European Commission.
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