Behaviors of Cementitious Materials in Long Term Storage and Disposal

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Behaviors of Cementitious Materials in Long Term
Storage and Disposal
Research Coordination Meeting on CRP Behaviour
of Cementitious Materials in Multipurpose
Packaging for Transportation, Long Term Storage
and Disposal , 24-28 Nov 08, Bucharest, Romania

• Chief Scientific Investigator I. Plecaš
• Additional Scientific Staff S. Dimovic, I.
  Smiciklas, D. Kicevic
• Vinca Institute of Nuclear Sciences,
  Belgrade-Vinca, Serbia
• Radiation and Environmental Protection Laboratory

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Institute of Nuclear Sciences "Vinca"Belgrade,
Serbia

• Development of Solidification Techniques
• for Radioactive Sludge
• Produced by a Research Reactor
• ILIJA PLECAS,PhD

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Abstract The reports present the results on
removal of sludge from the bottom of the spent
fuel storage pool in RA reactor, mechanical
filtration of the pool water, sludge
immobilization by cement , conditioning and
storage.
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The processing of radioactive wastes may be done
for economic reasons (e.g. to reduce the volume
for storage or disposal, or to recover a
"resource" from the waste), or safety reasons
(e.g. converting the waste to a more "stable"
form, such as one that will contain the
radionuclide inventory for a long time).
Typically processing involves reducing the volume
of the waste solidifying non-solid wastes to make
them physically stable, and packaging the waste
to isolate it from the environment.
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The RA research reactor, (6.5 MW) was shut down
in 1984 in order to reconstruct and improve all
vital reactor systems. However, for a number of
political, administrative, economical and
technical reasons, this reconstruction has never
been completed. SNF in the storage water pools
inside the RA building is leaking !!!
To solve the problems, Government decide to
propose VIND Project, (VINCA INSTITUTE NUCLEAR
DECOMMISSIONING PROGRAM) Three Sub Projects 1.
Spent Fuel Transport 2. Decommissioning of RA
Reactor 3. Radioactive Waste Management at
the Vinca site
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RA reactor SNF room
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The spent fuel storage pool (Fig. 1a) on the RA
research reactor in the Vinca Institute
consists essentially of four, six meters deep,
inter-connected rectangular basins.
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Pool water radioactivity
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Sludge sedimentation in the pool
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Presently, water in the reactor RA- spent fuel
storage pool is in very bad condition. Water in
the pool is dirty and its chemical parameters are
not maintained to minimize corrosion process.
Following the recommendations obtained from
(IAEA) the Vinca Institute elaborated a project
incorporating the following steps preliminary
removal of sludge and other debris from the
bottom of the pool in RA reactor, washing of
deposits from all the surfaces in contact with
the pool water, venting of the aluminum barrels,
mechanical filtration of the pool water, final
removal of the sludge, sludge conditioning and
storage at the waste repository at the Vinca
Institute site. Newer hangar has an space for
radwaste materials storing only for 1-2 years.
So, attempts are made in the Vinca in
developing the immobilization process for
conditioning low and intermediate level
radioactive waste materials and their safe
disposal into the appropriate disposal system.
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The cementation, as an immobilization process,
for the certain radwaste materials origin and
composition is investigated. Developed
immobilization processes have, as a final goal,
production of the solidified radwaste matrix
mixture form, that is easy for handling and that
satisfies safety and QA requirements, according
to radionuclide inventory, decay heat, radiation
dose rate and contamination, identification,
configuration and weight, mechanical integrity
for interim storage and the final disposal of
such materials on the appropriate sites. Radwaste
materials that were immobilized in the inactive
matrixes are to be placed into the concrete
containers, for the further management and
disposal.
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2. Characteristics and quantities of radioactive
sludge in the spent fuel element storage pool In
order to estimate storage conditions for the
spent fuel elements in the storage pool and
characteristics and quantities of the sludge on
the bottom of the pool, water and sludge samples
have been taken out from different locations in
the pool. Analysis of the water from the pool (pH
8.4, electrical conductivity 446 µS/cm, Cl
66 mg/L, Cu 0.05 mg/L, Zn lt 0.01 mg/L,
Fe 0.15 mg/L, SO4 55 mg/L) shows that
the water is highly corrosive to aluminum alloys
. Activity concentration of the water from the
pool of about 80 - 90 kBq/L of 137Cs nuclide,
although not of grave concern, is certainly
significant, and is incontestable proof that some
amount of the fission products is leaking.
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3. Sludge conditioning and storage Total quantity
of sludge on the bottom of the RA research
reactor spent fuel storage pool was estimated to
be about 3 m3. Estimation was made on the basis
of the average sludge height on the bottom of the
pool and pool surface. The sludge color has been
a dark red - brown, like an iron oxide corrosion
products. Gamma spectrometry analysis showed that
the specific activity of the sludge is about 1.8
0.2 MBq/L from 137Cs nuclide and about 15 kBq/L
from 60Co nuclide
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Cascs Based on the previous experience, a
technology was developed for sludge
immobilization and conditioning in a cement
matrix, inside casks, produced using the standard
200 L metal barrels which have lids supplied with
screws. Casks have been produced as concrete
shielded containers in standard metal barrels.
Thickness of the concrete walls is from 8 to 10
cm. Entire inner side of the cylindrical concrete
wall is covered by plastic tube with wall
thickness of 1 cm, which has been used as a model
in forming cylindrical concrete wall. This
plastic tube serves as a first barrier in
preventing radionuclides leaching from
radioactive sludge immobilized in a cement
matrix. The bottom cask concrete wall is also 6
7 cm thick. In order to prevent or reduce
radionuclide leaching, this wall has been covered
with epoxy resin. The useful volume of such
designed cask is about 75 L.
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The existing pilot cement mixer was reconstructed
to enable placing a barrel containing the planned
quantity of sludge on its platform without a risk
of spilling. About 60 65 l of sludge are poured
at a time from the sedimentation vessel into a
previously prepared cask. As soon as a cask is
filled up, it is hermetically covered with a lid
supplied with screw and transported to the
laboratory for sludge conditioning. There,
additional settling of sludge is allowed.
Separated water is pumped into a plastic can and
taken back to the RA reactor spent fuel storage
pool. Through the second stage of the sludge
settling, volume of the sludge in the cask has
been reduced to about 40 l. Fig. 2. A new
mechanical manipulator, which provides mixing of
the cement matrix with the sludge in the entire
volume of the barrel, was constructed.
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Fig. 3. Modified pilot mixer with concrete
container made in metal barrel.
Fig. 2. Apparatus for sludge settling
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Casks have been produced as concrete shielded
containers in standard metal barrels. Thickness
of the concrete walls was 10 cm.
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Well experience
Dr Ilija Plecaš, and my team published more than
40 papers in International Journals and more than
90 papers on International Conferences in the
field od Radioactive Waste Management, especialy,
on the problemas of Immobilization of Rad. Waste
in liquid and solid forms with cement.
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2. RADIONUCLIDE MIGRATION THROUGH POROUS
MATERIALS The dispersion of radionuclides in
porous materials, such as grout or concrete, is
described using a one dimensional differential
model.(Burns,1971, Lu,1978,Moriyama,1977) .

where KF - retardation factor ()1 D -
diffusion coefficient (cm2/d) or (cm2/s) A -
concentration in liquid (mol/l) or (Bq) X -
length (cm) Vv - velocity of leachant fluid
(cm/d) f - porosity ()1 ?T - bulk density
(g/cm3) kd - distribution coefficient (ml/g) t
- time variable (d). Using Laplace
transformation method, Eq.(1') becomes
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from which we can calculate a retardation factor,
KF. The coefficient of distribution, kd, can be
calculated
in which Vv, X, ?T, t and Ao are known. An
and De can be determined experimentally using a
leaching test procedure.(Hespe,1971) . For the
interpretation of the results of leach tests
shown in the following figures and tables, leach
coefficient D, is used, and it is defined as
where D - leach coefficient (diffusion) (cm2/d)
or (cm2/s) m - (?An/Ao)?(1/??t), slope of the
straight line (d-1/2) Ao - initial sample
activity at time zero (Bq) (Table I) An -
activity leached out of sample after leaching
time t, (Bq) t - duration of leaching renewal
period (d) (1,2,3,4,5,6,7,15,30,60) V - sample
volume (cm3) S - sample surface (cm2).
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Table I. Representative formulations of concrete
compositions as grams, for 1000 cm3 of concrete.
The cement specimens were prepared with a
standard Portland cement .
Portland Cement Sand 0-2 mm Aggregate 2-4 mm Aggregate 4-8 mm Aggregate 8-15 mm Water Aditive
Sample 1 400 672 85 463 724 150 8
Sample 2 400 692 75 423 794 150 8
Sample 3 400 822 91 595 476 150 8
Sample 4 400 662 73 317 933 150 8
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Table II Leach coefficients De(cm2/d) in
different concrete samples after 60 days, using
Eq.(4)
Leach coeff. Formula Formula Formula Formula
Leach coeff. C1 C2 C3 C4
De, 60Co 6,20?10-6 5,20?10-6 3,10?10-6 2,20?10-6
De, 137Cs 4,52?10-5 2,10?10-5 6,70?10-6 7,30?10-5
Table III Retardation factor KF and coefficients
of distribution kd(ml/g), after 60 days, ?T2,5
(g/cm3). f0,15-0,30
Coeffic. Formula Formula Formula Formula
Coeffic. C1 C2 C3 C4
KF, 60Co 98,0 99,6 104,2 105,3
kd, 60Co 6,9-16,8 7,0-17,1 7,3-17,9 7,4-18,1
KF, 137Cs 22,7 24,5 98,4 15,5
kd, 137Cs 1,7-3,8 3,8-6,7 6,9-16,9 7,7-18,3
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Compressive strength M(MPa) of samples after 28
days
Sample 1 40,0
Sample 2 34,0
Sample 3 37,0
Sample 4 38,0
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About 60 65 L of sludge are poured at a time
from the sedimentation vessel into such a
previously prepared cask. As soon as the cask is
filled up, it is hermetically covered with a lid
supplied with screw and transported to the
laboratory for sludge conditioning. When the cask
with the settled sludge is placed on the platform
of the mixer for further conditioning, the
necessary amount of cement, according to the
established formula of cement matrix and the
cement-sludge ratio, are poured into the cask.
The formula of cement matrix and the cement to
sludge ratio, are defined in accordance with
previous experience and experimental
investigations on radwaste cementation and and
experiments made with this sludge. The best
sludge to cement mass ratio for appropriate
mechanical strength was approximately 11.8
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Homogeneous substance could be obtained by less
than one hour mixing. This technology for sludge
conditioning eliminates all contamination and
radiation risks related to pouring the sludge
into the concrete mixer and pouring the
cement-sludge mixture into the metal barrel. The
barrel with the homogenized mixture is removed
from the mixer platform and placed in a separate
room for concrete to harden. It was
experimentally determined that the time needed
for concrete hardening is about 48 h. Due to
dynamics of the of the sludge removal from the
spent fuel storage pool, and sludge settling, the
final stage of radioactive waste conditioning has
took place 7 10 days after the sludge
cementation.
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Taking into account the measured sludge activity
concentration, two stage sedimentation process
(the first one in the vessel for sedimentation
and the second in the concrete cask container)
and the conditioning technology, it is estimated
that each cask with conditioned sludge contains
about 150 - 200 MBq from 137Cs nuclide and about
7 10 MBq from 60Co nuclide, i.e., specific
volume activity of the conditioned radioactive
waste in radioactive waste packages is about 0.7
- 1 GBq/m3 from 137Cs nuclide and about 35 - 50
MBq/m3 from 60Co nuclide. Taking into account
composition of radioactive waste packages, the
effect of self-absorption in homogeneously
dispersed radioisotopes in the cement matrix, and
concrete cask walls radiation absorption
capability, the contact gamma-ray dose rates,
measured on the casks surface, were in the range
from 0.1 to 0.15 mSv/h, i.e., much less than 2
mSv/h, which is an acceptable limit value for the
radioactive waste packages.
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Practically, Reduction Factor, RF , was about 33
( 200m3 of liquid radioactive waste from RA
basins, we transferred in 31 concrete casks in
metal barrels, each 200 lit., approximately 6 m3
of solidified sludge with concrete protection).
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31 concrete casks in metal barrels are disposed
in hangar H2, at the existing radwaste repository
at Vinca Institute site.
31 concrete casks
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4. Conclusion After the operations, explained
above, have been performed, necessary elements
for planning further stages of pool and water
cleaning and treatment of the spent fuel should
be obtained. Through many years research and
development in radioactive waste immobilization
and conditioning performed experimental
experience gave the possibility to choose the
best formulation for cement mixture and results
gave us certainty to claim that described methods
and used matrix materials will serve as a
barriers to preserve radionuclides migration to
the surroundings for at least 300 years.
Optimization of the processes and matrix-radwaste
mixtures is in further progress and we hope that
this work will influence the design of the future
Serbian storage center, shallow land burial type
for low and intermediate level radioactive
wastes. All performed steps have been done in
accordance with all relevant requirements for
radiation safety and radiation protection .
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