Class 19
Nuclear Spent Fuel Reprocessing
1
Introduction
• During its time in the reactor, the fuel (bars, tubes…) is
subject to important physical and composition
modifications due to the neutrons irradiation:
• The fissile material content (U-235 or Pu-239, Pu-241)
decreases progressively by fission
• U-239 generates by capture Pu-239, which disappears by
capture or by fission.
– The capture reactions lead to the formation of Pu-240, Pu-241, Pu242 (with Pu-241 fissile). This apparition of new fissile isotopes
compensate only partially the diminution of the global content of
fissile matter
• Apparition of new elements in the fuel, baneful to the chain
reaction progress
– Transuranic elements (Np, Am, Cm)
– Fission Products (Sr, Cs, Tc…) some of them are neutrons poison
such as Gd
2
Introduction (2)
• The modifications of the fuel composition
associated to the strong heat release by
fission provoke important changes in the
physical state of the fuel.
• Crystals structure modifications (holes or
concentrations of atoms)
• Variation of the volume:
– The volume occupied by the atoms created by
fission is greater than the one of the
disappeared matter.
– Moreover, some fission products are gaseous
and their solubility in uranium is practically
3
non-existent
Introduction (3)
• All of these changes will alter the physical
properties and the structure of the fuel with
modifications of the thermal, mechanical,
dimensional characteristics.
• Consequently the cladding can be
deteriorated, which can go to the formation
of cracks or even break.
4
Introduction (4)
• The following implies that, after a certain
period of irradiation time, it is necessary to
take out the fuel from the reactor.
– Decrease of the content of fissile material
– Progressive poisoning of the fuel
– Risk of cladding break
• This operation is performed before the
consumption of all initial fissile material
5
Introduction (5)
• The reprocessing activity answers 2
objectives:
– Recovery of the fissile material (U-239, Pu-239,
Pu-241) to reuse it for the fabrication of new fuel
elements (example Recycling Pu via MOX fuel
fabrication, recycling of U from PWR for new
enrichment)
– Separation of nuclear waste (activation and fission
products) as a function of their pollution in order to
store then to avoid any potential danger and
release towards the biosphere
6
Irradiated Fuels Characteristics
• The irradiated fuels are taken out of the
nuclear reactor after a certain time that
can vary : (PWR: 3 years, fast breeder
reactor: 2 years)
• The following table is presenting the
principal characteristics of some nuclear
fuels
7
Principal characteristics of
irradiated fuels
Reactor type
Form
Nature
Cladding
Combustion Rate
(Mwd/t)
UNGG
Bars
U metal slightly
alloyed
Magnesium
3,000 to 5,000
Heavy water
Rods
Natural UO2
Light water
Fast neutrons
Submarine
reactors
Plates
assembly
Zircaloy
10,000 to 15,000
Enriched UO2
(< 5%)
(U,Pu)O2 with
15 to 20%
of Pu
Zircaloy,stainless
steel
Stainless steel
20,000 to 40,000
50,000 to 100,000
U-Al
U-Zr
(enriched U up
to 93%
Aluminum
Zircaloy
30%
< 30%
8
Irradiated Fuels Characteristics (2)
• If we consider only the characteristics
important to the reprocessing, we can
distinguish:
– The composition (metal, alloy, oxide, carbide…)
– The enrichment of the fissile material density
– The structure or the form (bars or tubes, rods,
plates, spheres..)
– Combustion rate whose depend the activity a, b,
g and the residual calorific power after irradiation
9
Irradiated Fuels Characteristics (3)
• For the “reprocessor”, the irradiated fuel can be
considered as a mixture of 5 families of
compounds:
– The fissile material (U-235, Pu-239 and Pu-241), which
represents the noble part of the fuel and whose
recovery is the main goal of reprocessing
– The fertile material (U-238, Pu-240)
– Heavy isotopes, they are neither fissile nor fertile (U236, Np-237, Pu-242, Am and Cm)
– The fission products (principal source of b and g activity)
– The other metals (Mg, Al, Zr, metals coming from
stainless steel) part of the fuel or forming the cladding
10
Irradiated Fuels Characteristics (4)
• The composition of one ton of irradiated fuel
coming from the following reactors
– UNGG
– PWR
– Fast neutrons Reactors
• is reproduced in the next table and is illustrated for
the PWR case in the following slide
• The content of fission products and of plutonium
increase with the fuel rate as
UNGG > PWR > Fast BR
11
Reactor Type
1200MW
Form and
Initial
Composition
UNGG
U metallic Bar
alloyed with
Mo or Al
Mg cladding
PWR
Rod containing
UO2 enriched
(3.25%)
Zircaloy Cladding
4,500 Ci
= 20 W per kg of fuel
1,800 Ci
= 75 W per kg of fuel
Fuel
characteristics
from nuclear
reactors
Total FP
Actinides
(g/ton)
Quantity of principal FP (g/ton
1,400 Ci
= 6 W per kg of fuel
Rod containing
(U and Pu)O2
Stainless Steel
Cladding
80,000 MWd/t
for the fuel
in the core reactor
Average Burn up
In the reactor
b and g activity
Residual power
After 150 days
cooling
Quantity
To be treated
Fast
Neutrons reactors
Total An
12
Characteristics of fuel irradiated during 3 years in a PWR
Initial Fuel (1,000 kg)
Irradiated Fuel (1,000 kg)
Fission Products
Different Isotopes
Of Pu (9 kg)
3
years
13
• The repartition in mass of the fission
products can be deduced from the curves:
Yield of fission vs. mass number of FP
• See next slide
14
Yield in mass of the fission (%)
Distribution curves of the FP of U-235
Fast Neutrons (14 MeV)
Thermal Neutrons
Mass Number of the Fission Product
15
Yield in mass of the fission (%)
Distribution curves of the FP of U-233, U-235,
Pu-239 (Thermal Neutrons)
Mass Number of the Fission Product
16
Irradiated Fuels Characteristics (5)
• An important characteristic of irradiated
fuels is their residual calorific power,
consequence of the activities b and g of
fission products and activation products
• It decreases over time, as presented in the
next table
17
Residuaire Power of irradiated
Fuels
Residual Power in w/kg after
Reactor
Type
Fuel Rate
MWd/t
90
120 days 150 days
days
180 days
UNGG
4,000
8.7
7.0
5.8
4.9
PWR
30,000
30
24
20
17
FBR
50,000 to
100,000
80
110
65
90
55
80
47
72
18
Irradiated Fuels Characteristics (6)
• The reprocessing of these different types
of fuels is not so much different,
nevertheless, on needs to take into
account the fissile material density, the
content of fission products, the calorific
power produced, the structure of the fuel
elements.
19
The chemical Treatment
Objectives and constraints
• The problem is as follows:
• Obtain separately and with a high percentage yield
Uranium and Plutonium decontaminated from the
fission products, in order to manipulate them later
as if they were materials that have never been
irradiated
• The decontamination factor of U and Pu in FP is
between 107 to 108
• The purification of U and Pu towards non
radioactive elements need to be also very specific
20
The chemical Treatment (2)
• The specifications to respect for finished
products U and Pu issued from
reprocessing of LWR and fast neutrons
reactors fuels are described in the next
table
21
Constraints for recovery and purification of finished products
EXTRACTION YIELD ≥ 99.5%
Decontamination Factors for Uranium and Plutonium (FP)
Cooled LWR Fuels (3 years)
b, g Activity ~ 7.5 105 Ci/ton U+Pu
aActivity ~ 105 Ci/ton U+Pu
``-Np 385g/ton U+Pu
-Pu 10,000 g/ton U + Pu
LWR
7,500 dpm = 5 mg Np
7,500 dpm = 10-8 g Pu
Fast Neutrons
7,500 dpm = 5 mg Np
7,500 dpm = 10-8 g Pu
FINISHED PRODUCTS SPECIFICATIONS
b, g Activity
~ 0.5 mCi/g U
~ 1 to 2 mCi/g Pu
aActivity
~ 1,500 to 15,000 dpm/g U
Cooled Fast Neutrons Reactor Fuels (6 m)
b, g Activity ~ 5 106 Ci/ton U+Pu
aActivity ~ 9.105 Ci/ton U+Pu
(100,000 MWD/t)
-Np 385g/ton U+Pu
-Pu 10,000 g/ton U + Pu
LWR PLANT
DF U
b,g = 1.5 .106
Pu = 2.105
Np = 100 (77)
DF Pu
b,g = 107 to 7.107
U ~ 6.4.105
(150 ppm in Pu)
FBR PLANT
DF U
b,g = 107
Pu = 1.3.107
Np = 36 (or 360)
DF Pu
b,g = 4.107
U ~ 1.5.105
Np 3 to 4
22
• Constraints to respect during reprocessing
operations are numerous. Indeed,
• The fuel is inside a water proof cladding, in
general refractory to the usual chemical
reagents
• The intense radioactivity implies very special
work conditions
• The fissile material masses require to be
very careful towards criticality risks
• The radioactivity of the effluents release into
the environment need to be very small and
lower than the limits established by agencies
such as EPA
23
• The next tables show the performances to
reach for the liquid and gaseous effluents.
The purification factors, ratio between
quantity that enters the plant and quantity
that exits the plant (via effluents) are
depending on the plant but are in general
very high, specially for alpha emitters.
24
Respect of the environment - Liquid Effluents
ANNUAL ACTIVITY TREATED
In UP2 800 + UP 3 (1600 t/year – LWR)
SITE: La Hague, France
b,g : 1.2 109 Ci/year
3H : 1.12 106 Ci/year
a : 1.6 108 Ci/year
RELEASE ASKED FOR La Hague
b, g : 45,000 Ci/year
3H : 5.3 106 Ci/year
a : 90 Ci/year
Purity Factor*
for the most
radioactive
Effluents
La Hague
Pur. Factor b, g = 2.6 104
Pur. Factor a = 2.6 104
The purification factors, ratio between quantity that enters the plant
and quantity that exits the plant (via effluents)
NB: American norms for liquid release
25
a = 0.5 10-3 Ci per GWe and per year for a PWR cycle.
This norm implies a purity factor of 1010 for the high active liquid effluents
Respect of the environment Gaseous Effluents
ANNUAL ACTIVITY TREATED
In UP2 800 + UP 3 (1600 t/year – LWR)
SITE: La Hague, France
Kr-85 = 1.7 107 Ci/year
3H =
1.1 105 Ci/year
I-129 = 61 Ci/year
Others FP = 109 Ci/year
ATMOSPHERIC RELEASE
ASKED FOR La Hague
Kr-85 : 107 Ci/year
3H : 5 104 Ci/year
I-129 and I-131 : 2Ci/year
Others FP = 10 Ci/year
Purity Factor*
for the most
radioactive
Effluents
La Hague
Pur. Factor Kr = 1
Pur. Factor I > 30
Pur. Factor a, b, g = 108
NB: American norms for gasrelease
Kr-85 = 50,000 Ci: I = 5.10-3 Ci per GWe and per year for LWR
which implies a Pur. Factor Iode > 200 and
Pur. Factor for Kr = 10 for La Hague
26
Principle
• The principle of chemical reprocessing of
nuclear fuels relies essentially on liquidliquid extraction.
• This choice implies the dissolution of
irradiated fuels in an aqueous solution, after
elimination of cladding material, followed by
the realization of the liquid-liquid extraction
cycles that leads to the chemical separation
– Uranium + Plutonium towards fission products
and other metals
– Uranium from Plutonium
27
• The universal process used today is the
PUREX Process.
• It uses a nitric dissolution of the fuels
• A specific separation of U and Pu by
extraction, using a solvent n-Tributylphosphate diluted in an aliphatic
diluent (dodecane).
• The scheme of the PUREX Process is
reproduced next slide
28
Fuel
Assemblies
GAS TREATMENT
Off-Gas
Treatment
Unloading
Iodine
Kr-Xe
WASTES
VENTILATION
Gaseous effluents
Storage
Pu
Dissolution
Clarification
shearing
Oxides
Storage Pool
1st cycle
TBP
Extraction
U(VI)-Pu(IV)
2nd
cycle
3rd
cycle
2nd
cycle
3rd
cycle
PuO2
Pu
nitrate
U
nitrate
UO3
U
SOLID WASTES
HLW
FP, Np, Cm, Am
Solid compounds
Concentration
Denitration
Interim
Storage
under water
UF6
UF4
UO2
Liquid
Effluents
RE-INTRODUCTION
IN THE FUEL CYCLE
TREATMENT
Interim liquid storage
Vitrification
Interim Storage of glass blocks in well
SLUDGE
Liquid effluents
29
The different steps of the Process
Deactivation and transport to the plant
• The first destination of the irradiated fuel is close
to the reactor
• This is the storage pool where the assemblies are
stored under a few meters of water
• This period of cooling allow an important fraction
of the radioactivity to cool down
• After a few months, the most instable
radionuclides, whose the half life time is in minute,
hour, day, have practically disappeared and there
are only the fission products left with long half-life
time.
• It is then unnecessary to wait any longer, the
irradiated fuel can be transported towards the
reprocessing plant.
30
Transport
• The fuel assemblies consisting of rod bundles
measuring about 4 m long, holding usually 264
rods (PWR 17*17) initially containing UO2 based on
uranium enriched to about 3.5% U-235, irradiated
to about 2.85*1012 J/kg (33,000MWd/t) for PWR
fuel, ( and about 2.42*1012 J/kg (28,000 MWD/t for
BWR fuels) are transported in shielded casks from
the power plant sites to the reprocessing plants.
• These casks, designed for the simultaneous
transport of several fuel assemblies are very heavy
(about 100 t) and complex machines that must
guarantee transport safety.
31
Transport (2)
• On their arrival at the reprocessing plant, the fuels
are unloaded form the transport casks.
• This delicate operation is normally performed after
the loaded cask is immersed in water
• To simplify this unloading operation, AREVA has
successfully installed a dry unloading facility at La
Hague, an operation designed to shorten unloading
time and to minimize the volume of contaminated
effluents to be treated.
• After unloading, the fuel assemblies are stored
under water in pools awaiting reprocessing
32
Spent fuel storage capacity
• The spent oxide fuel storage capacity pf
the reprocessing pants is already
considerable.
• As an example, La Hague has 5 pools with
a total capacity of 10300 t currently in
operation.
33
The different steps of the Process (2)
Decladding and dissolution clarification
• First decladding takes place
• Decladding can be performed
– Chemically if the material of the cladding and
the form of the fuel allow it (case of the
magnesium cladding for UNGG)
– Mechanically for the UNGG, PWR and fast
neutrons reactors fuels
34
Decladding operations
(La Hague, Windscale …
Rod end
Cylinder
Thumb
Wheel
35
Shearing
36
Shearing
• To enable nitric acid solution to attack the fuel, it is
necessary to chop the assemblies.
• This operation is performed at the reprocessing
plant, after cutting the top and bottom ends of the
assemblies, using a horizontal shear which
accommodates the complete bundle.
• Each rod of the fuel element is thus broken into
pieces about 30X 35 10-3 m long, which contain all
or part of the nuclear material.
• At this stage a fraction of the gaseous fission
products escapes through the dissolver off gas
where it is subjected to iodine trapping by NaOH
scrubbing and absorption on silver loaded inorganic
solid sorbent.
37
The different steps of the Process (3)
• Once the decladding, and shearing are
performed, the fuel dissolution can take place.
• It is performed on pieces of the cladding and
the fuel by mixing to a nitric acid solution
• The dissolution reactions are:
UNGG
U
9
( H   NO3- ) (UO22  ,2 NO3- )  1.55NO  0.85NO2  0.05N 2  2.25H 2 O
2
PWR and
Fast Neutrons and
Reactors
3UO2  8( H   NO3- )  3(UO22  ,2 NO3- )  2 NO  4H 2O
PuO2  4( H   NO3- )  3( Pu 4  ,4 NO3- )  2H 2O
38
Dissolution
39
Dissolution (2)
• Small amounts of residues remain during the
dissolution phase of PWR and fast neutrons
reactor fuels.
• These residues are made of small part of
cladding and polymetallic inclusions
containing fission products: Mo, Ru, Rh, Te,
Pd + U, Pu whose the quantity increases with
the combustion rate.
40
Dissolution (3)
• Fuel dissolution produces a solution with
the following approximate composition:
• U(VI) = 250 kg/m3 (g/L)
• Pu(IV) = 2.5 kg/m3 (g/L)
• FP = 9 kg/m3 (g/L)
• The b and g activities are approximately
200 Ci
• The a activity is approximately 1.87 Ci/L
41
Dissolution (4)
• The fuel is not totally dissolved and about
3 kg of FP is not dissolved.
• These are the highly active dissolution
fines consisting of FP: Pd, Rh, Ru and Mo.
• A small fraction of Pu is found trapped in
these residues.
42
Dissolution (5)
• Implementation of the dissolution operations
must take into account the criticality risks.
• Tow alternatives are available:
• Poisoning of the dissolution liquor by a neutron
absorbent (such as gadolinium) and
• Criticality safety guaranteed by the geometry.
• The next figure presents 2 types of continuous
dissolver, (rotary dissolver UP3, La Hague and
helicoidal continuous dissolver (TOR,
Cadarache).
43
Wheel = 4m
Metal Feed
engine
Continuous
Rotary
Dissolver
Discharge of
Insolubles
engine
Suspension with spring
Dissolution
Solution
Nitric Acid Feed
Feed
Solid
Wastes
Fast Annular
Continuous
Dissolver
Dissolution
Solution
44
Clarification
• Before being sent to the U and Pu
extraction cycles, the dissolution liquors
must be clarified to remove any traces of
solid particles, whose presence would be
detrimental to the extraction operations.
• The solution is cleared by centrifugation or
filtration and adjusted to 4M HNO3 before
going towards the extraction unit.
45
• The gas emitted from the dissolver contain
nitrous vapors NOx and certain fission
products, the noble gas Krypton and Xenon
as well as Iodine (volatile as I2 form).
• Iodine is trapped as stable form PbI2 and
the Nox gas are recycled as nitric acid,
reaction performed by water wash,
countercurrent of the gaseous flow plus
addition of hydrogen peroxide:
H2O
NOx + O2
HNO3
46
Extraction Cycle
• The nitric acid dissolution solution of the irradiated fuel
contains the totality of uranium and plutonium as nitrate
complexes (UO2 2+, 2NO3-) and (Pu 4+, 4 NO3-) and the
fission products.
• The extractant chosen, tri-butyl phosphate (TBP) is
diluted to 30% by volume in an aliphatic hydrocarbon, ndodecane, HPT or a petroleum cut.
• TBP presents a great affinity towards U(VI) and Pu(IV).
• Its affinity towards the FP and other metals is low, which
brings a very selective extraction of U (VI) and Pu(IV)
47
• During the contact between the organic
phase and the aqueous phase, the 2
following reactions take place
(UO22  ,2 NO3- )  2TBP  UO2 ( NO3 ) 2 (TBP ) 2
( Pu 4  ,4 NO3- )  2TBP  Pu( NO3 )4(TBP ) 2
• The reactions are strongly displaced
towards right (extraction) if [NO3-] is high
and towards the left (reextraction) if [NO3-]
is low
• The choice of HNO3 allows to predict the
direction of the solutes (U(VI) and Pu(IV))
transfer
48
1st extraction cycle
• The 1st decontamination cycle (U + Pu) is
presented on the next slide.
• The aqueous flow (U + Pu + FP) is contacted with
the organic phase in a pulsed column,
countercurrent flow.
• The first stage is hence a co-decontamination
stage during which the U and Pu are removed
from most of the FP
• The organic phase charged in U and Pu is then
washed by a nitric acid solution in order to scrub
from the organic phase the small quantities of FP
that was entrained with U and Pu into the organic
phase.
49
1st extraction cycle (2)
• After this first extraction column, most of the PF are
separated from U and Pu and exit as raffinate.
• U and Pu exit the column as organic phase
• The organic phase is stripped of U and Pu, and the
solvent can be recycled after treatment, to eliminate the
degradation products H2MBP (monobutyl phosphate) and
HDBP (dibutyl phosphate), formed by the hydrolysis and
radiolysis of TBP.
• U and Pu prepurified exit the second column in the
aqueous solution.
• In certain units, the implementation of the codecontamination cycle may be complicated by the
behavior of certain FP such as Zr and Tc.
50
HNO3
H2O
Solvent
Extraction
U + Pu
Solution
U + Pu + FP
FP
1st EXTRACTION CYCLE
Reextraction U + Pu
U + Pu
Solvent
Wash
Solvent
Treatment
U + Pu
51
1st extraction cycle (3)
Solvent
U+Pu+PF*
+
Solvent
Aqueous
Solution
U+Pu+ FP*
Feed
U+Pu
+
Solvent
PF*
Mixing
Settling
52
Solvent Extraction
Mixer settler
Solvent
U+Pu+PF*
+
Solvent
Aqueous
Solution
U+Pu+ FP*
Feed
U+Pu
+
Solvent
Feed
Aqueous Phase
U+Pu+FP*
U+Pu+FP* Solvent
U+Pu
PF*
Solvent
Out
Loaded with
U+Pu
Solvent
In
PF*
Exit
raffinate FP*
Mixing
Settling
Centrifugal Contactor
Pulsed Column
Solvent
In
Moteur
Feed
Aqueous Phase
U+Pu+FP*
* FP : Fission Products
Solvent Out
Loaded U +
Pu
Exit
raffinate FP*
Protection
lead / concrete
Feed
Aqueous Phase
U + Pu + FP*
Interphase
Exit Solvent
Loaded U + Pu
High Mixer
Settler
Compressed Air
Perforated
Plates
Exit
raffinate
FP*
Feed Solvent
Low Mixer settler
53
2nd Extraction Cycle- Splitting Stage
• The second step consists in separating U/Pu realized by
liquid-liquid extraction (see next figure).
• It exploits the different redox properties of U and Pu.
• In the first extraction column, the co-extraction U(VI)/Pu(IV)
is again performed.
• After washing, the organic phase loaded in U(VI)/Pu(IV) is
injected into the second column where it is contacted with
an aqueous solution containing a specific Pu reducer.
• U/Pu splitting takes place by reductive stripping of the
plutonium.
• Indeed, Pu(IV) is reduced to Pu(III) and migrates into the
aqueous phase, the organic phase containing U(VI) is
treated by water (countercurrent) to realize the extraction
of U(VI) into the aqueous phase.
• From this step, U and Pu will go in separated purification
cycles (2 more cycles)
54
Effluents
SEPARATION U/Pu CYCLE
Pu
H2O
Solvent
Treatment
Reextraction U
Aqueous
Solution Wash
Solvent
Extraction
U + Pu
Solution
U + Pu
Reextraction
Pu
Reducer
U + Pu
Solvent
Wash
HNO3
U
55
Reducer for Pu
Several Pu reducer can be used
• Ferrous sulfamate (Fe 2+, 2NH2, SO3-)
Pu 4+ + Fe 2+
Pu 3+ + Fe 3+
• Uranium (IV)*
2Pu 4+ + U 4+ + 2H2O
2Pu 3+ + UO2 2+ + 4 H+
• Hydroxylamine nitrate (NH3OH+, NO3-)*
2 Pu 4   8 NO3-  NH 3OH  , NO3-  2 Pu 3 
(*: the most used)
1
1
N 2O  H 2O  3H   9 NO32
2
56
Final conditioning
Uranium
• A purified uranyl nitrate solution is concentrated
by distillation and the concentrated solutions ( U
> 400 to 500 g/L) are essentially stored waiting
to be reused.
• For U coming from PWR, a fraction is recycled
for new enrichment, which requires the steps of
precipitation/filtration of ADU, thermal
decomposition which gives U3O8 and then
transformation into UF6
57
Final conditioning (2)
• Uranium (2)
• The recycling of uranium produced by LWR fuel
reprocessing is to as easy as the U-235 content
of 1.07% would tend to imply.
• In fact, the uranium thus produced is also
enriched with isotopes of atomic weight 232
(harmful due to hard g emitters decay products)
and neutron absorbents U-234 and U-236.
• In France, however, the decision has been taken
to recycle reprocessed uranium
58
Final conditioning (3)
Plutonium
• Plutonium is precipitated from the purified
aqueous solution by addition of oxalic acid
Pu 4   2H 2C2O4 2 Pu(C2O4 ) 2 - 6H 2O  4H 
HO
• The plutonium oxalate, after filtration is
calcinated at 450°C, which gives PuO2
450C
Pu(C2O4 ) 2 .6H 2O 
 PuO2  6H 2O  4CO2
59
Final conditioning (4)
Plutonium(2)
• The plutonium recycle strategy currently adopted
by many countries consists in the preparation of
mixed fuels (U, Pu)O2 (~7% Pu) called MOX for
LWR reactors.
• A plant is build in France, MELOX, with a
production capacity of 100 tons of heavy metal
per year.
• A plant is being constructed in USA (August
2007), the MFFF facility, Savannah River Site
(see class 20)
60
Final conditioning (5)
Military Plutonium
• In the case of military application of
plutonium, Pu metal will be fabricated, 2
steps are necessary:
Fluorination
500C
PuO2  4 HF  
 PuF4  2 H 2O
High
Calciothermy
PuF4  2CaTemperature
 Pu  2CaF2
61
Waste Conditioning
• Most of the radioactivity of the irradiated
fuels can be found in the waste issued from
the reprocessing plant.
• These waste from different flows (gas,
liquid, solid) need to go towards different
treatment units for
– Reduction of their volume
– Immobilization in specific matrices for their
temporary or permanent storage to avoid their
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release towards the biosphere
Gaseous Waste
• The different gaseous flows from the decladding unit and
the off-gas from the dissolvers contain dangerous
radionuclides: Kr-85, I-131(I2) and Ru-104 (RuO4)
• Furthermore considerable quantities of NOx gas are
produced, that need to be recycled. Gas are washed,
countercurrent wash, with water in the presence of H2O2,
which gives the formation of HNO3 that can be recycled.
• NaOH trapping will trap I2 and Ru, radioactive iodine can
be conditioned as PbI2, that can be safely stored.
• After going through different filters, the gas, that contain
still radioactive isotopes of noble gas such as Kr-85, are
rejected into the environment through chimneys that allow
the rapid active gas dilution.
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Liquid Waste
• The essential of FP is in the raffinate of the
1st extraction cycle.
• The other radioactive effluents produced
by the other purification cycles contain
lower amount of radioactivity.
• A second series of liquid effluents is made
of degraded organic phases
• We will see in class 21 the treatment of
these effluents
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Solid Waste
2 categories
• Waste issued from the process
–
–
–
–
Cladding
Gas Traps (PbI2, AgI)
Muds produced during the treatment of liquid effluents
Solid products issued form the final treatment of the
solvent
– Ion exchange resins ( treatment of the pools)
• Technological waste
– Apparatus
– Pumps…
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Solid Waste (2)
• The main solid residue issued from
dissolution consist of sections of clad or
hulls weighing 300 kg per ton of fuel.
• For the time being, these wastes are
stored at La Hague under water,
amounting to interim storage. In the UP 3
and UP2 800, La Hague France, the hulls
are immobilized in a concrete matrix for
subsequent final disposal.
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Solid Waste (3)
• The different waste are classified as a
function of the radioactive type and its
importance (a, b, g, low, medium and high
level waste).
• The conditioning is specific, it can be
• Cementation (PWR hulls)
• Wrapping in Epoxy resin (for the ion
exchange resins)
• Bitumen isolation (mud from the effluent
treatment unit)
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