Government of India
Bhabha Atomic Research Centre
Nuclear Recycle Group
STIPENDIARY TRAINING PROGRAMME
CAT - I
REPROCESSING OF NUCLEAR FUELS
BY SOLVENT EXTRACTION PROCESS BASED ON TBP
A. RAMANUJAM
Notes compiled by
Training & Special Assignments Section
PREFRE PLANT, TARAPUR
INTRODUCTION
The recovery and purification of fissionable material produced by neutron capture
in uranium is the principal reason for reprocessing of fuels irradiated in reactors. The
great majority of fuels is either all metallic or an oxide sheathed in metals. The nearly
universal approach to recovery of metallic or oxide nuclear fuel is by dissolution with an
inorganic acid, adjustment to suitable salting strength concentration, acidity and oxidation
state and then separation of the source and fissionable materials from each other and from
base metals and fission products by solvent extraction.
Processes other than the conventional aqueous methods include volatility and
pyro- metallurgical methods. Although these latter methods afford potential advantages
over the over the aqueous methods, they are most significant when combined with fuel
fabrication step into a closed fuel cycle. However newer methods are only in the research
or developmental stages and cannot be compared as proven processes. For the aqueous
processes, fuel dissolution and feed preparation vary widely depending on chemical
composition of fuel and cladding, but generally involve mechanical or chemical
decladding followed by dissolution of the fuel in Nitric acid.
The above mentioned
1. Decladding of fuel
2. Dissolution of fuel and
3. Feed adjustment and pretreatment if any,
are grouped as Head End Treatment of the fuels.
1
6
HEAD END TREATMENTS
A) DECLADDING OR DEJACKETING OF FUEL
The conventional nuclear fuels are clad in a jacket, which keeps the fuel and its fission
products from coming in contact with the heat transfer medium. Their general purpose is
(1)
(2)
(3)
(4)
To prevent erosion and corrosion of the fuel.
To act as a heat transfer medium with out undue fouling.
To entrap and contain radioactive fission product gases and
To restrict the growth of fuel during fission.
Aluminium Zirconium and Stainless Steel are popular jacketing materials although
others like Nichrome and Niobium and Ceramics may possibly be used in the future.
The removal of jacket material prior to processing is desirable because it contributes
greatly to the
volume of highly radioactive process waste that must be stored and
it increases the volume of process feed stream and
hence the size of the equipment for various process steps. The jacket removal may be
by mechanical or chemical means. The chemical means have been employed widely
t6odate. In some cases simultaneous or integral dissolution of jacket and fuel may be
advantageous as in the case of BUTEX PROCESS where when the aluminium cladding
is dissolved integrally with uranium , it will act as a "salting agent" to improve the
e….xtraction of uranium into methyl Iso butyl ketone.
(1) CHEMICAL DEJACKETING.
(A) ALUMINIUM - Caustic or caustic-nitrate dissolution :
Aluminium cladding or jacket is removed batch-wise from natural uranium slugs by
boiling with 5-10 wt % NaOH. Sodium nitrate is also added to give an off gas of NH3
rather than H2 as the latter gives an explosive reaction with O2. The NaOH concentration
beyond 10% gives a faster rate of all dissolution, which is difficult to control. The batch
reaction time is normally 8 to 10 hrs. The reaction for mole ratios of 1/1.65/1.47 for
AL/NaNO3 may be written as:
Al + 0.85 NaOH + 1.05 NaNO3 = NaAlO2 + 0.9 NaNO2 + 0.15NH3 + 0.2 H2O
(B) ZIRCONIUM
Zirconium either pure or alloyed with tin to form zircalloy is used as jacketing material
for fuel elements. Using H.F., Zr can be dissolved from the natural uranium
rods without appreciable loss of U or Pu.
The Zircex process involves the chlorination of zirconium at 300 to 6000 C with an
anhydrous HCl to form volatile ZrCl4.
6
2
H.2SO4 has also been used for zircalloy jacket removal.
(C) STAINLESS STEEL
The 18-Cr-8Ni-type stainless steel such as types 304 and 347 can be dissolved in
excess boiling 4-6 M H2SO4 (Sulfex process) or in mixed HCl-HNO3. They have not
been studied in detailed manner. The mixed HCl-HNO3 acids are not suitable for jacket
removal since uranium losses would be high. Carbonization of stainless steel by methane
at 10000 C converts the alloy to a HNO3 soluble form although passivation may occur.
Electrolysis gives rapid and complete dissolution of stainless steel in HCl but large-scale
studies are still to be done.
The rate of dissolution of S.S. type 304 and 347 in boiling 400 to 600% excess 4-6
H2SO4 is about 11 mg/cm2/min. The solubility of UO2 is refluxing 6 M H2SO4 is low. No
data are available on highly irradiated fuels of any kind.
(62) MECHANICAL DEJACKETING:
Of the many dejacketing methods studied rolling milling extrusion, abrasive cutting,
electrical discharge cutting and shearing are effective for unbonded or Na-K bonded fuels.
Only rolling and shearing appear applicable to metallurgical bonded fuels.
For further details, refer to Reactor Handbook.
……………
6
3
6.
DISSOLUTION
The most common dissolvent for uranium is boiling nitric acid. The net reaction:
U + 4.5 HNO3 =
UO2(NO3)2 + 1.57 NO + 0.84 NO2 + 0.005 N2O + 0.043 N2 + 2.25 H2O
(1)
In practice, a reflux condenser is placed on the dissolver and oxygen or air added to
the vessel, which allows considerable recovery of oxides of nitrogen as reusable acid. The
nitric acid consumption is ultimately reduced to 2.0 moles per mole of uranium as given
by the reaction
U + 2 HNO3 + 3/2 O2 = UO2 (NO3)2 + H2O
(2)
The off gas consumption varies with the initial concentration of acid. The
dissolution rate is accelerated by the presence of dissolved uranium. Uranium dioxide
dissolves in nitric acid by the net reaction.
UO2 + 4 H + + 2 NO2 = UO2++ +2NO2 +2H2O
(3)
Th6e solubility of Uranyl nitrate in water is relatively high, solutions of 2M in
UO2++ are commonly used, and the solubility of Uranyl nitrate varies with nitric acid
concentration. It is possible to dissolve uranium with any net evolution of gaseous
pr.oducts except the fission products according to equation 2 which uses oxygen as
reactant and this type of dissolution is called Fumeless dissolution.
(A) FEED PRETREATMENT
Solutions resulting from the dissolution of spent fuels must be adjusted in most cases
before it is ready for solvent extraction. The most common treatment involves one or
more of the following operations
(1)
(2)
(3)
(4)
(5)
Solids removal or stabilization.
Adjustment of feed acidity
Adjustment of salting strength by dilution or evaporation
Reduction oxidation of uranium or plutonium and
Scavenging and pretreatment of certain undesirable fission products.
(1) SOLIDS REMOVAL OR STABILIZATION: The purpose of solids removal is
necessary to avoid plugging and emulsification in solvent extraction equipment.
Insoluble residues in dissolution may result from oxide coatings on fuel moderator,
coolant residues jacket bonding residues, catalyst residues (such as Mercury) or some
insoluble components of fuel. These solids may be separated by various mechanical
means.
(2) FEED ACIDITY: The acidity of an extraction column feed may be important in
.avoi6ding reactions with solvent, in obtaining the conditions for reduction oxidation
4
f various ions, in establishing optimum conditions for salting the desired components
to avoid excess acid in stripping sections or to furnish the salting agents. This step
will be discussed further at a later stage.
(3) REDUCTION OXIDATION: In PUREX PROCESS at the end of dissolution,
uranium is present mostly in sixth valence state and plutonium in tetravalent state.
The Plutonium valence state is further adjusted to tetravalent state by the addition of
NaNO2 followed digestion for 2 hours at 500 C. The addition of NaNO2 may not be
always necessary since at high radiation levels nitrate is converted to nitrite.
PuO2++ + NO2 - + 2H+ = Pu+4 + NO3- + H2O
6.Pu+3 + 2NO2 + 8H+ = 6Pu+4 + 4H2O +N2
The dual function nitrite (i.e. it reduces Pu (VI) to Pu (IV) and oxidizes Pu (III) to
Pu (IV) makes it an ideal feed conditioning agent in PUREX process where Pu (IV) is
most extractable form in TBP. NO2 gas also can be used for this adjustment. The redox
reactions will be discussed further during later stages.
(B) CONCENTRATION AND SALTING STRENGTH
In many dissolution it is possible to obtain a sufficiently high concentration of
products in solution so that further concentration by evaporation is unnecessary. If the
dissolver solution is relatively unsaturated it may be desirable to concentrate it as much as
possible to allow a high uranium feed concentration and there by obtain the maximum
processing rate for a given size of extraction equipment. This topic will again be
discussed further during next chapter.
(C) SCAVENGING AND OTHER PRETREATMENT
Solutions resulting from the dissolution of spent fuels may be treated chemically and
/ or physically to obtain partial decontamination of fission products and to remove
undesirable solids prior to introduction of dissolver solutions to the solvent extraction
cycle.
(a) Volatilization of Iodine and Ruthenium and their removal during dissolution by
sparging
In. fuel reprocessing, sparging refers to passage of an externally supplied gas through
the dissolver solution to remove volatile fission products by vaporization. Iodine and
R6uthenium are volatilized during and following dissolution and scrubbed from the
off gases to prevent entry into atmosphere. This partial removal of these fission
products by sparging provides additional decontamination supplement that obtained
in the solvent extraction cycles and reduces the number of cycles required for desired
product purity to minimum.
(b) SCAVENGING
5
In t.he application of solvent extraction to fuel recovery it has been found that certain
fission products are particularly difficult to separate from the product. Thus in Redox
process ruthenium is the principal fission product contaminant of uranium and in PUREX
process Zr and Nb are the most difficult fission products to separate Head end processes
have been developed for the removal or scavenging of specific fission product from the
dissolver solutions. In the case of Zirconium and Nb, similar extraction behavior to
uranium in TBP and a tendency to form radio colloid and carry over with the solvent
characterize their chemistry. Ruthenium is the most difficult fission product to eliminate
because of its many valence states. Ru in dissolver solution generally constitutes from 5
to 10% of the total gamma radioactivity present depending on the length of the time
between the end of irradiation period and the start of chemical separations. If no specific
steps are taken to remove Ru before the start of solvent extraction, the proportion of Ru in
the intermediate solvent extraction cycle products rise from the 5 - 10% value dissolver
solution to 45 - 75% and continues to increase in subsequent cycles, eventually reaching
as high as 85 - 100% in each of the final products. Consequently it is desirable to remove
as much Ru as possible in the Head end step.
Ru can be removed from the dissolver solutions of uranium fuel by volatilization as
RuO4 using a strong oxidant such as Ozone. KMNO4 and Con.HNO3. The principal
reactions are:
6
3Ru+4 + 4M NO-4 + 4H2O = 3RuO4 (gas) + 4MNO2 + 8H+
Ru4+ + 2 O3 + 2H2O
= RuO4 (gas) + 2 O2 + H+
Ozone treatment will remove 90% of Ru.KMNO4 allows for removal of Ru as
RUO4 as well as the carrying of Zr and Nb on the resulting MNO2. 90 % or more of Ru,
Nb and Zr fission products are removed in this process. The handling of RuO4 has proven
difficult because of its tendency to reduce to RuO2 and collect in pipelines and on the
surface of the oxidizer vessels. When adsorbed in a caustic scrubber RuO4 ends up as
suspension of RuO2.
Variables in the permanganate oxidation of ruthenium include MNO4 loss by acid
catalysis increasing MNO4 loss with temperature and the presence of reducing agents like
Cr3, which increases the KMNO4 decomposition.
Zr and Nb scavenging on MNO2 was the result of studies on MNO4 oxidation of Ru
The MNO2 formed from the decomposition of MNO4 is very finely divided and a highly
specific adsorbent for Zr and Nb species. The MNO2 is then separated from the system by
centrifugation or filtration. Then the MNO2 is washed free of U and Pu and dissolved
with a reducing agent before disposal.
6
PUREX PROCESS
6
INTRODUCTION
Essentially all plant scale processes currently used for separating plutonium from
irradiated uranium are based on solvent extraction. These processes which yield
plutonium and uranium highly purified with respect to fission products are vastly superior
to the original precipitation processes which they have replaced. Solvent extraction
processing of plutonium and uranium is dominated by the PUREX process employing a
Tributyl Phosphate - hydrocarbon solution as extractant.
6
Generally the reprocessing of fuel rods by the PUREX process consists of following
steps: (1) Decladding or dejacketing
(2) Dissolution
(3) Feed preparation
(4) Co-decontamination cycle.
(5) Partition Cycle.
(6) Final U purification cycle.
(7) Pu reconversion.
The first three steps together are called head end treatment and have been discussed
previously.
The aim of the fourth step or the co-decontamination cycle is to separate uranium and
plutonium together from the rest of the fission product impurities.
The aim of partition cycle is to separate and purify uranium and plutonium from each
other and from fission products.
6
The aim of the sixth and seventh step is to purify U and Pu further individually before
conversion to metallic form.
The extent of purification or decontamination of uranium or plutonium achieved in
each step is generally expressed by a term called decontamination factor or simply DF
and is defined by the following expression:
Beta or Gamma activity/gm of U or Pu in feed
DF =
Beta or gamma activity/gm of U or Pu in product.
As the steps 4,5 and 6 are essentially solvent extraction cycles, the solvent
extraction process will be discussed in a detailed manner before a discussion on these
7
process steps is taken up. Larger the DF greater is the purification and efficiency of the
process.
DF after 2 cycles of extraction
U
Pu
Beta
6.6 X 106
7.7 X 106
Gamma
2.7 X 106
2.0 X 106
SOLVENT EXTRACTION (GENERAL PRINCIPLES):
8
The term solvent extraction implies the separation of one component from a
mixture by its preferential solubility in an extracting solvent. The extraction capability of
a solvent for a particular solute is expressed by the term equilibrium constant.
C org.
Equilibrium constant Ea* = K =
C Aq.
Where C org. is concentration of the solute in the organic phase and C Aq. is the
concentration of the solute in the aqueous phase at equilibrium. K is called
distribution coefficient or equilibrium constant. This ratio is constant
(a) for a pair of liquids
(b) for a particular solute
(c) at a definite temperature.
This ratio is also independent of
(a) initial concentration of the solute in the aqueous phase and
(b) quantities of liquid involved with in a limited range. This law is known as
partition law and is obeyed only when the solute exists in the same chemical form in both
phases, with out association, dissociation or polymerization.
So this Ea* or K value is an index of the extent and ease with which a solute can be
extracted by an organic solvent.
Greater the value of K greater is the extraction. If the K values of two solutes or a
particular solvent under a particular set of conditions vary very much, it is implied that
these conditions one of the solutes can be preferentially extracted by the organic liquid .
This is expressed by a term called separation factor.
F =
KA
------KB
Where , KA is dist. Coe. of A
KB is dist. Coe. of B
F is separation factor
Since there is always an equilibrium between the concentration of the solute in the
organic phase and aqueous phase, unless K value is extremely large and it is not possible
to extract completely and quantitatively all the solute from the aqueous phase into organic
phase in a single contact. In a single contact only that much amount permitted by K value
will be taken up by a definite amount of solvent
100 X D
% E = ----------------D + (VA/VO)
Where D is the dist.Coe.
VA is volume of aqueous phase
VO is volume of organic phase.
For a quantitative removal of a particular solute from the aqueous phase, the
organic phase after first extraction must be removed and the aqueous phase must be
brought in contact with a fresh batch of organic solvent. Thus by repeated contacts the
solute can be quantitatively removed into organic phase. For example, if in a system a
solute has K = 10 and assuming empirically 110 units of solute were originally present in
the aqueous phase: 9
On a single contact:
90 units of solute will go to organic phase leaving behind in the aqueous phase 10
units. By any amount of shaking 10 of the solute cannot be taken up by the solvent
On a second contact:
9.1 units of the solute will go to organic phase and 0.9 units of the solute will remain
in the aqueous phase.
On a third contact:
0.82 units of the solute will be extracted by the solvent and 0.08 units will still remain
in the aqueous phase. Thus in three contacts, if K value = 10, 99.93%of the solute is
removed from the aqueous phase and in each contact a finite amount of the solute remains
in the aqueous phase. If the K value had been 0.1 instead of 10, it can be worked out that
after 3 extractions only 25% of the solute goes into organic phase. This is called BATCH
process. This method is followed in laboratories. However, on industrial scale in solvent
extraction columns what is followed is continuous counter current extraction process. In
a continuous counter current extraction the aqueous phase is injected at the top of a
column and flows downwards, the organic liquid is injected at the bottom of the column
and because of its lower density it flows upwards, thus the aqueous phase as it travels
down the column meets fresh batch of solvent in stages and the solute is being
continuously extracted into the solvent. This in principle is equivalent to shaking the
aqueous phase with several batches of organic solvent. The aqueous phase when it
reaches the bottom of the column will be deplete of the solute (called raffinate) and the
organic solvent floating at the top of the column will be pregnant with the solute (loaded
solvent). The two phases are then pumped out separately.
Therefore, in such a process, separations are also dependent on the relative flow ratios
of the two phases and the extent of contact between the two phases. Because of this, in
such process K values can be lower than in batch process and still the separation can be
efficient (K value of 1 - 10 is considered very good). As mentioned earlier because of
the finite value of K, if the loaded organic solvent is brought in contact with a fresh batch
of aqueous phase containing no solute part of the solute that has gone to the organic
solvent will come back to aqueous phase. This referred to as stripping. By repeated
contact with fresh batches of aqueous phase all the solute in the organic phase can be
stripped into aqueous phase. But if stripping is done under the same conditions of
extraction where K is larger (tendency for the solute is to go to organic phase) only very
small amount of solute will be stripped into aqueous phase and so too larger a number of
stripping will be required to have a quantitative removal of the solute into aqueous phase.
So for stripping, conditions are so chosen, that the K value of the system is extremely low
(tendency for the solute is to go to aqueous phase).
K values of different solutes under any set of conditions will not be same. This
preferential distribution and stripping of one or few elements forms the basis of solvent
extraction separations.
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SOLVENT EXTRACTION CYCLES
A solvent extraction cycle consists of all operations involved in preferentially
extracting these solutes back into aqueous phase until the organic phase is virtually free
from the solute once more.
SEPARATION OF TWO SOLUTES:
Consider two solutes A & B, A is essentially inextricable in some organic solvent and
B is extractable. The aqueous feed stream containing A & B are fed to solvent extraction
column are contractor at a point near the centre.
Aq. Scrub Solution
Loaded Organic
Aq. Feed Solution
Org. Strip Solution
Aqueous Raffinate
Fig. 1
As the aqueous stream flows down towards the bottom of the cascade nearly all of
the solute B and a small amount of A are extracted into organic phase, which is flowing
towards the top of the cascade. Most of the solute transfer takes place in the section of
the column between the feed point and the bottom of the column and the section is
therefore called extraction section. As the organic passes the feed point, it is met by a
fresh aqueous stream that serves to wash or scrub the A (and little of B) from the organic
phase back into aqueous phase. This section of the column above the feed point is
therefore called scrub section. A single column or contactor having both these sections is
called a compound contactor.
A compound contactor is necessary for the separation of two components unless the
ratio of their distribution coefficients is extremely large. The function of the extraction
section is as far as possible to place all B into organic phase and the function of scrubbing
section is to place all A into aqueous phase. The solute concentration of B changes but
little in the scrubbing section especially in the organic phase but it falls of rapidly in the
extraction because of its extraction into and removal with organic stream. In a like
manner, the concentration of A changes but little in the extraction section especially in the
aqueous phase, but it falls of rapidly above the feed point.
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The solvent extraction cycle is completed by stripping or back extraction of
component B into aqueous stream in a separate contactor called stripping column.
SEPARATION OF THREE OR MORE SOLUTES:
Consider a mixture of three solutes A, B & C of which A is essentially inextractable
but B & C can be readily extracted into organic phase. The mixture is fed near the centre
of a compound contactor and the inextractable solute A leaves the contactor in the
raffinate and other two solutes leave in the organic. The inextractable solute A is
essentially completely removed from the solvent in the scrubbing section of the first
contactor. The products B and C in the loaded solvent are subsequently separated in
another center fed contactor. In this contactor, B is stripped from the organic phase in the
top half by a fresh aqueous phase and any C that is stripped along with B is washed out of
the aqueous phase by fresh organic solvent in the bottom half. The contactor in which B
and C are separated is called partitioning contactor. Finally, organic phase is stripped of C
in a separate contactor.
Aq. Strip agent only for B
Solvent C
Org. Feed B & C
Org. Scrub to extract C
Aq. Product B
Fig. 2
In the above-mentioned topic, A consists of all Fission Products and B & C stand
for Plutonium and Uranium respectively.
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CO-DECONTAMINATION CYCLE
SELECTION OF TBP AS AN EXTRACTANT FOR THE EXTRACTION OF FISSILE
AND FERTILE MATERIALS (PU & U):
Generally following factors influence the choice of a solvent for the extraction process
in the processing of irradiated fuel elements. A final choice involves a compromise
between various factors.
(1) Highest possible extraction or partition ratio for Pu and U.
(2) Extraction selectivity.
Extraction selectivity should be positive for Pu and negative for fission products. (If
individual constituents of a mixture of elements present in an aqueous solution are to be
separated the solvent must preferably extract one or as few elements as possible under a
particular set of conditions, or if more than one element is extracted, the extent of
extraction of one should at least vary very much from that of the other. Only then, a good
separation can be achieved. This is referred to as selectivity and is expressed by the
difference in log K values of two elements. S = log KA - log KB should be large if A
and B are to be separated. The selectivity is affected by (a) solute concentrations (b)
presence of other elements only such a solvent which has high selectivity by adjusting the
above mentioned factors can be chosen.
(3) Ability to re-extract the components back into aqueous phase easily.
(4) Chemical stability
It should be resistant to attacks by HNO3, HCl etc. to a reasonable extent.
(5) Thermal stability:
It should remain stable to operational temperature of 700C. but below 1200 C.
.
(6) Radiation stability
It should be resistant to radiation damage at radiation levels normally encountered in
processing of irradiated fuels under consideration.
(7) Low mutual solubility
Even if the extraction is efficient, if the solvent has considerable solubility in the
aqueous solution, there will be a loss of solvent during each operation of the cycle.
Therefore, the solvent must have very low solubility in the aqueous solution and also the
dissolution of water in the organic phase should be minimum.
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(8) Density difference between two phases:
Density is the factor deciding the ease with which organic phase separates from the
aqueous phase. For an efficient extraction, the two phases must separate easily and
quickly. This would be so, only if the difference between the densities of the two layers is
sufficiently large. Normally the solvent should have a low-density compared to aqueous
phase.
(9) Low viscosity:
The solvent should have low viscosity to facilitate pumping and other mechanical
problems associated with the handling of the solvent.
(10) Toxicity:
As far as possible it should have low toxicity and its vapour should not be dangerous.
(11) Vapour pressure:
The solvent must have low vapour pressure (or high boiling point) so that it will not be
inflammable. Combined with low volatility this will prevent fire hazards and toxic
problems.
(12) Cost and recoverability:
The solvent must be obtained in large quantities as a commercial product at low price
and should be purified by easy and simple means. Simple chemical treatments should
suffice for the recovery of used solvent.
SOLVENTS GENERALLY USED IN EXTRACTION PROCESSES.
Different solvents that meet the above specifications to a reasonable extent are:
(1) Methyl Iso Butyl Ketone (Hexone)
(2) Dibutyl Carbitol (DBC)
(3) Tributyl Phosphate (TBP)
(4) Trilauryl Amine (TLA)
and few others.
Of these, TBP is generally favored and is employed in most of the fuel reprocessing
plants in the world.
 This is highly selective of U and Pu.
 Since this is used as a plasticizer in the industry it is available cheaply as a
commercial product and can be purified easily by treatment with NaOH and
HNO3 to remove any decomposition products.
 This has a high boiling point (2660 C.) and
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









non-volatile. It exists in liquid form from -780 C. Onwards.
The flash point is 2940 C. F (Cleveland open cup).
Its dielectric constant is 7.97 (at 300 C.). It does not have a definite melting point
at least down to -1960 C. low temperature TBP passes on to a glass state.
Its solubility n water is 0.4 g/l and
water solubility in TBP is 64g/l.
a resists being attacked even by Conc.HNO3.
It has high thermal chemical and radiation stability as would be required. It has
very low extractability for fission products and so the purification of U and Pu is
excellent only two unfavorable properties are
(a) high-density 0.973 and
(b) high viscosity 33.2 millipoise.
Both of these can be compensated by diluting it with an inert diluent like
kerosene. Even a 95% diluted TBP has high extractive power and so these
disadvantages are not serious.
THE MECHANISM OF ACTINIDE AND FISSION PRODUCT EXTRACTION BY
TBP.
The main object of the process may be envisaged as the recovery of uranium and
plutonium separately or together for further use and the segregation of fission products in
smallest volume for safe disposal. Solvent extraction process accomplishes this in a fuel
reprocessing plant by taking advantage of the varying extent to which the three
constituents present from nitrate complexes which are extracted by TBP.
For UO2++, PuO2++, Pu4+ and for few other ions in tetravalent state nitrate complexing
is considerable. But for few exceptions like Ru+3, Zr+4 the ability to form nitrate
complexes is very poor for most of the elements that constitute the fission products. On
this basis three main groups can be formed.
Group Metal or Radical Nature of the complex
Group
A
Strong
B
Weak
C
Very Weak
Metal or Radical
UO2+2, PuO2+2
Pu(IV), U(IV),
Zr(IV), Ce(IV),
Ru(NO)III
Pu(III), Y(III),
Ce(III), La(III)
Pr(III), REs, Nb(V)
Cs+,Sr+,Ba+
Ru(IV), Rh(IV)
Mo(IV), Tc(IV)
Nature of Complex
Tetra or Hexa Nitrate
complex
Di or Tri nitrato complex
Trinitrato complex
Doubtful
Because of the similarity of U (IV) and Pu (IV) in the formation and extractability of
nitrato complexes and almost inextractability of fission products an efficient separation of
former from the later is attained in a single step in solvent extraction with TBP. Then
taking advantage of the fact that Pu (III) is weakly extracted a further separation of U
15
and Pu is achieved by reducing Pu (IV) to Pu (III) in which state it gets stripped from
the organic phase. TBP being a nonionized solvent (Dielectric constant 8) mechanism of
extraction is U and Pu nitrates associate in the aqueous phase readily solvated by TBP and
are taken into the organic solvent.
Species extracted by TBP:
Each nitrate occurs in one form only in TBP as the neutral m unionized molecule,
unhydrated but solvated by a definite number of solvent molecules. If M is the metal
atom, the species extracted are:
Trivalent state
Tetravalent state
Hexavelent state
M(NO3)3 3TBP
M(NO3)4 2TBP
MO2(NO3)2 2TBP
In all cases the 6 coordination positions of the central metal atom is filled up.
EXTRACTION OF NITRIC ACID:
The important mechanism for the extraction of nitric acid is
H+ + NO3- + TBP <=====> HNO3 + TBP
The extraction of nitric acid is almost independent of temperature except at low
concentrations of HNO3. The salting agents that do not themselves extract such as Na or
Al nitrates increase the extraction of nitric substantially at low acidity. At higher acidity
the effect is not much pronounced. Uranyl nitrate, thorium nitrate and other extractable
salts, Back extract nitric acid from TBP (i.e. HNO3 is pushed out to aqueous phase).
Extraction of Uranium and Plutonium and the parameters affecting the extraction:
TBP is bound to metal ions by covalent bonds but to acids and water by hydrogen
bonds. The extraction mechanism of U (IV) into TBP is well established and the reaction
at equilibrium is
UO2++ + 2NO3- + 2 TBP <====> UO2 (NO3)2 2 TBP
(UO2(NO3)2 2 TBP)
Ku = ----------------------------(UO2++) (NO3-)2 (TBP)2
C org
Dist. Coe: Kd = ----C Aq
U Org.
= -----U Aq
.
. . Kd = Ku (NO3-)2 (TBP)2
16
or Kd (NO3-)2 (TBP)2
Similarly for plutonium the reaction is :
Pu+4 plus 4(NO3)- + 2TBP <====> Pu (NO3)4 2 TBP
(Pu(NO3)4 2TBP)
K Pu = ------------------------(Pu) (NO3)4 (TBP)2
Kd
COrg
Pu Org
= ------ = ---------CAq
Pu AQ
.
. .
Kd (NO3)4 (TBP)2
C4H9O
NO3 NO3
OC4H9
|
\ /
|
C4H9O - P --------> O ----->Pu<--->O <-------- P - OC4H9
|
/ \
|
C4H9O
NO3 NO3
OC4H9
17
PARAMETERS AFFECTING THE EXTRACTION OF PU AND U:
(1) TBP concentration :
As can be seen from the equations given in previous section the extraction (Kd) of Pu
and U is directly proportional to the square of TBP (free) concentration at any time.
This proportionality however exists only upto 10% TBP beyond which it decreases
slightly.
(2) Acid concentration:
As can be seen from the derivations generally the extraction coefficiency of Pu (IV) is
directly proportional to the fourth power of NO3- concentration under normal conditions.
Similarly, the uranium extraction power is proportional to square of the nitrate
concentration in the aqueous phase.
When the nitrates of uranium and plutonium are extracted by TBP from varying
concentrations of nitric acid solutions, Kd values increase sharply with aqueous nitric acid
concentration, passes through a maximum and falls a little thereafter. The initial rise is
due to increase in the NO3- concentration that acts as a salting agent by common ion
effect and enhances the formation of undissociated neutral species that are readily
extracted. However, at high acidities the HNO3 itself competes with U and Pu and
therefore there is a fall in the Kd value.
(3) Salting agents:
Some common ion effect can be introduced by adding neutral salt, which is not soluble
in the organic phase into the aqueous phase along with uranium and Pu (like Al (NO3) 3
or NaNO3). In this way nitrate ion concentration can be increased without increasing the
nitric acid concentration. Because of this the K value Increases even at low acidities.
Normally salts of cations with higher +ve charge are better salting agents. But the acidity
of the aqueous phase should not be reduced too low as plutonium has a tendency to
hydrolyze at low acidities and hydrolyzed species cannot be extracted. Generally salting
out agent is avoided in the extraction procedures followed in the fuel reprocessing plants
as the concentration of raffinate (the aqueous phase remaining after extracting U and Pu
containing all fission products) cannot be done beyond a particular limit where the added
salts separate out. This puts a limit to the volume to which the radioactive wastes can be
reduced.
(4) Solute concentration:
18
It is observed that there is a difference in the extractability when trace quantities are
present and when "Macro " amounts are present. The effects are two types (a) When
macro amounts are present, there is an increase in (NO3) concentration than when trace
quantities are extracted. This factor is known as self-salting and it increases the
extraction. (b) When large amounts of solute are present there is a greater use of
available TBP and so TBP gets saturated. This retards further extraction as the saturation
effect sets a limit to the capacity of TBP for any given element.
Theoretically (for e.g.) 30% TBP in shell sol T (1 M) can take up uranium to the
extent of 119 g/L (0.5 M) only, as per the formula UO2 (NO3) 2 2TBP. This is called
100% saturation. In certain cases, another effect of saturation will be the limited solubility
of the solvated complex itself in the diluent. As a result the organic phase separates out
into two layers. This phenomena is called the third phase formation which is a
complicating factor in a column operation and should be avoided, by controlling the
solute concentration, acidity, temperature or changing the diluent.
(5) Valence states:
The extractability of U and Pu of different valance states:
III little extracted
IV more strongly extracted KPu(IV) > KU(IV)
VI very strongly extracted KUO2+2 > KPuO2+2
(6) Temperature effect:
The temperature effect on the Pu and U extraction by TBP is complex one which
depends on U,H+ and TBP concentration but the increase or decrease in Kd is
marginal due to change of temperature. This fact is utilized in some column operations
in PUREX PROCESS and is called dual temperature process.
19
CHEMISTRY OF TROUBLESOME
EXTRACTION BEHAVIOR WITH TBP:
FISSION
PRODUCTS
AND
THEIR
RUTHENIUM:
The hypothesis of a multiplicity of species has been considered for a long time as the
explanation of behavior of ruthenium in TBP. Such a proposition could easily account for
the fact that the bulk of ruthenium is relatively easy to separate but that a small part
remains with the products in spite of drastic scrubbing of the organic phase. A
complicated complex ion chemistry combined with slow equilibria is characteristic of
many transition elements. The proposal is the RuNO3+, nitrozyl ruthenium and the
complex ions formed from this entity are the ruthenium troublemakers. Nitrosyl
ruthenium is formed via ruthenium (IV) during the dissolution & also subsequently in a
manner that is not yet entirely clear. A large fraction of ruthenium is nitrosyl ruthenium
by the time it reaches solvent extraction contactors. The nitrosyl ruthenium is a very
stable entity and resists the most vigorous attack, even oxidation to RuO4 and distillation.
The nitrosyl ruthenium has five coordination positions available to be filled in aqueous
nitrate solutions by water, nitrate, the nitro group or the hydroxyl ion. The nitro
complexes are quite stable and are in the majority. The complex
RuNO(NO2)2(H2O)2(OH) considered to be the dominant nitro complex. It is extracted
well by ethers and Ketones but very fortunately only moderately by TBP. The nitrate
nitrosyl ruthenium complexes are probably the specific troublemakers in PUREX
processes. In acidic solution the following equilibria are involved.
RuNO (H2O)53+ + NO3- ==> RuNO(H2O)4(NO3)2+ + H2O
RuNO (H2O)4(NO3)2+ + NO5- ==> RuNO(H2O)3(NO3)2+ + H2O
RuNO (H2O)(NO3)2+ + NO3- ==> RuNO(H2O)2(NO3)3 + H2O
There is some evidence that further substitution of nitrate ion for water occurs at
high concentrations of nitric acid. The extractability of various species by TBP is
Ru(NO)(H2O)2(NO3)3 > RuNO(H2O)3(NO3)2+ > RuNO(H2O)4(NO3)2+
> RuNO(H2O)53+ (inextractable).
It is considered that TBP does not displace water molecules from the primary
coordination sphere during extractions but it is associated in some less definite manner.
After the extraction has taken place TBP replaces water in the primary sphere. For
20
example RuNO(TBP)2(NO3)3 is formed from RuNO(H2O)2(NO3)3. Also all the equilibria
given above for aqueous phase are believed to occur in organic phase also. The
Pu(NO)(TBP)2(NO3)3 has such large extraction coefficients (of the order of 1000)
that it cannot be scrubbed back from organic phase until it is converted to a less
extractable species. The conversions back and forth between species are shown in both
phases. So the trinitrato species and to a less extent the dinitrato species extract in
extraction section of the compound contactor. As soon as the extraction occurs the
substitution of TBP for water in the coordination sphere of Ru commences and highly
extracted species start to form. When the organic phase passes in to the scrub section, the
dinitrato species is rather readily scrubbed out and the trinitrato species less readily.
Finally the TBP substituted nitrato complexes scrub out only slowly as they are converted
to water substituted form. This slowly conversion continues in stripping contactors and
releases Ru into the product streams.
ZIRCONIUM:
From what is known of Zr chemistry it should contrary to experience have
straightforward behaviour in TBP processes. In practice, great part of Zr does follow just
a simple behaviour but a small fraction is extracted and scrubbed out very slowly only.
No detailed explanation of the behaviour of this abnormal zirconium has been reported. It
is this small fraction which so often limits the overall DF. The solutions from dissolved
fuels contain some colloidal material, which in combination with zirconium extracts well
and in the scrubbing Zr is released very slowly.
NIOBIUM:
The process behaviour of Nb similar to Zr "Normal” Nb extracts very little but a small
fraction follows the Zr pattern. No explanation has yet been found for this behaviour.
21
CHOICE OF DILUENTS
The diluent for TBP must be non-polar unreactive toward HNO3 and HNO2 and
stable to radiation. In addition it should have a high flash point and low vapour pressure.
These criteria limit the choice of hydrocarbons containing 12 to 14 carbon atoms and
since extraction is similar among these diluents a choice is often based primarily on
resistance to HNO3 and HNO2. Among hydrocarbons straight chain hydrocarbons
(paraffins) are more resistant than branched chain paraffins and both types are much more
stable than napphthenes, olefines or aromatics. Generally n-dodecane or n-tetra decane or
the mixture of the two (1800 - 2100 F) are probably the best diluents. This cost however
has prevented its widespread use in plant scale processes. High-grade kerosene with low
aromatic and olefinic content has proved to be satisfactory compromise between desirable
properties and cost and has been extensively used. Somewhat superior to kerosene, are
Shell-Sol. T, Sol trol etc. which are also used widely.
CONCENTRATION OF TBP IN DILUENT FOR PROCESS CYCLES
A concentration of 30 volume percent of TBP in suitable diluent has been prominent in
all PUREX PROCESS works. The choice of TBP concentration is a compromise between
processing a minimum liquid volume and having a solvent phase with suitable physical
characteristics to be processed. for ex., in small mixer settlers it becomes almost
impossible to operate the unit with 45% TBP loaded with U . The chief difficulty is that
in the transient conditions during start up the organic phase extracts so much U that it
becomes the denser phase and blocks the aqueous flow. The maximum allowable
concentration of TBP must depend on the contactor design. For that reason 30% TBP is
again not a unique flow sheet condition but only a representative one.
DEGRADATION OF TBP
TBP is very stable as organic compounds go but is subject to hydrolysis or dealkylation giving rise to DBP(R-O)2P-OH), MBP and butyl alcohol. Nitric acid reacts
directly to give DBP and butyl nitrate. The formation DBP is only the first step of
complete de alkylation that leads ultimately to orthophosporic acid. This hydrolysis can
take place in both phases. If tetra butyl phosphate is present it gives DBP at a faster rate.
if the solvent is held for protracted periods at 500 to 600 C, it can degrade considerably.
At temperature in excess of about 1200 C TBP can be decomposed with explosive
violence in presence of uranylnitrate and nitric acid as oxidant. In addition to gaseous
products butyl nitrates, butyl ether and butyric acid are formed along with a variety of
unidentified materials.
22
The major practical consequence of radiolysis of TBP is the production of DBP2. Also
at high radiation doses significant amounts of polymeric materials are formed that cause
emulsion and bind U very tightly. These polymeric materials have some of the
characteristics of long chain phosphoric and phosphoric acids. The radiolysis of TBP also
Produces a broad spectrum of chemicals ranging from hydrogen gas to high molecular
weight hydrocarbons in extreme cases. The production of DBP is reflected most directly
and most crucially in adversely affecting decontamination (primarily of Zr). A very crude
estimate of the effect is:
(a) When the solvent receives 0.1 watt/hr. of radiation per liter in the contactor process
performance is unimpaired.
(b) When the solvent receives one-watt hour / L noticeable but not too serious effects are
observed.
(c) When the dose is 100 watt hr. / L there may be catastrophic loss in decontamination.
PUREX process normally falls in the first category.
The presence of DBP increases the extraction co efficiency of U, Pu, Zr and Pa. dBP
interferes with the separation of U and Pu and makes complete stripping of both elements
difficult. DBP can have drastically deleterious effects on decontamination from Zr. Zr
and Pu(IV) form DBP complexes of formulae M (DBP)4 which is highly extractable
in solvent. In partitioning columns iron forms ppt with DBP which carries Pu (III). Zr and
Pu (IV) form ppts. with PO4 if present which are highly insoluble in both phases in
usual PUREX system and drift about forming interface crud in the extraction column
which picks up lot of fission product activity, there by subjecting the TBP phase to
much higher dose than what normally would have been encountered otherwise. This in
turn would further aggravate the degradation of TBP.
However, DBP being acidic can be easily washed away with any basic solution and
the TBP is always washed with Na2CO3 and NaOH solution before being recycled.
DILUENT DEGRADATION
Only general remarks can be made about the degradation of diluent. Especially in the
case of degradation of kerosene, chemical attack by nitric acid and nitrous acid interacts
with radiolytic attack to produce a spectrum of nitrogen compounds, ketone esters and
unsaturates.
Chemical attack seems to consists of a combination of oxidation and nitration. Once
these have occurred in the intermediates are available for a complex of organic reactions
with nitrous acid playing a prominent role. Significant changes in chemical properties
usually occur well before important physical changes which may in extremely advanced
cases can go so far as to increase the viscosity and density. The general chemical effect is
to produce new extractants more potent than TBP, which may prevent complete stripping
of Pu and U in stripping contactors. But perhaps more important is their great affinity for
Zr which is so great that the extractants even if present to only 10 parts per billion can be
determined in semiquantitative way by the use of radio Zr tracer. Tracer Zr in HNO3 is
equilibrated with alkaline washed degraded solvent. The Zr extracted by TBP is washed
away easily with dilute HNO3 to leave the Zr that is bound tightly by the degradation
products. Using the specific activity of Zr, the concentration of Zr that is tightly bound
by diluent is known, which when expressed as moles per billion liters indicates the extent
to which the solvent has degraded and is given as Z no. of the solvent.
23
Radiation damage follows a complicated course, but the most important consequence
is to form olefins. The olefins are the starting material for the formation of the
troublemakers mentioned above. The unsaturated olefins are also direct troublemakers
because of their ability to react with fission product Iodine. This can lead to radioactive
Iodine being scattered up and down the extraction system.
These potent extractants cannot be easily washed away by acid or alkali.
CO - DECONTAMINATION CYCLE: EXTRACTION.
The aim of the cycle has already been explained as the separation of the U and Pu
together from fission products. After adjustment of the composition of dissolver it is fed
to the center of a contactor which might be a compound contactor. Nitrous acid is added
to the feed stream to ensure that PU is in IV state which is the most extractable form in
TBP, although this addition is not absolutely required always. A solvent stream 30% TBP
is in Shell-Sol T is fed at the bottom of contactor and a stream of 2 or 3 M scrub HNO3
enters the contactor from top. U and Pu are extracted together by solvent stream, the
scrub stream scrubs out most of the small quantity of fission products that has extracted.
The fission product leaves the contactor in the aqueous raffinate at the bottom and Pu and
U leave in the solvent stream. This is the general way in which the extraction contactor
of the co-decontamination cycle is operated.
During the operation a variety of variations are possible and the following parameters
are worth considering.
(1) Acidity
(2) Flow rate
(3) Other operational optimums
Generally speaking the flow sheet conditions for the operation of this extraction
column either belongs to high acid or low acid flow sheet.
Low acid flow sheet generally means extracting at 1 M HNO3 and scrubbing at 2 M
HNO3 as against 2 and 3 M in high acid flow sheet. When the extraction is carried out at
low acidity all fission products remain essentially in aqueous phase especially so in the
case of Zr as the E* of Zr increases with increasing acidity. As ruthenium extracts better
at low acidities, it is extracted to some extent but is scrubbed by the 2 M HNO3 high acid
scrub flowing from top. Hence, low acid flow sheet is expected to give a better
decontamination from Zr but at the expense of Ru decontamination as compared to high
acid flow sheet. In the case of high acid flow sheet Zr DF may be less compared to low
acid flow sheet. But not much difference is observed in DF at low temp., as compared to
high temperature operation where Zr refluxing happens. High acid flow sheet ensures a
better recovery of U and Pu and low acid flow sheet is not safe from Pu loss and some Pu
refluxing might occur. Theoretically more degradation of TBP can occur at high acid
operation in comparison with low acid operation. All in all it is doubtful that one type of
flow sheet is greatly to be preferred over the other.
24
The general operation technique is to load the organic phase to obtain a high saturation
value and to keep the acidity at such level as to extract all Pu into organic under these
conditions. The extraction faster for U than for Pu. It is essential to have a high loading
of U in solvent phase in order to reduce the extraction of fission product
(as the
available free TBP concentration is less after full loading). Too much loading of U into
organic on the other hand will prevent the extraction of Pu also. Therefore, an optimum
saturation of 60 to 85% is preferred generally. The scheme used generally is given below.
The interface during extraction is maintained at the bottom to avoid entry of raffinate into
product organic stream (column - organic continuous - aqueous disperse).
Aqueous Scrub.
3 M HNO3
0.3 V
Loaded Solvent Org.
PRODUCT
Scrub Section
Aqueous Feed
1 Vol. U 360 g/l
Pu 2.16 g/l
H+2 M
Extraction Section
O/A=3/1
Fresh Org. Feed
30% TBP
4 Vol.
Aqueous Raffinate.
Fig. 3
25
CO - DECONTAMINATION CYCLE :
STRIPPING
The loaded solvent from extraction column containing U and Pu is passed through a
separate scrub column if the F.P. content is high. Then the solvent is fed from the bottom
of stripping contactor. (The feed contains normally U = 80 g/L Pu = 350 mg/L H =
0.3N). A strip acid stream flows down from the top. Lowest acidity gives maximum
stripping of U and Pu as they have low Kd values at these acidities. But it should not be
lower than 0.01 M to avoid Pu hydrolysis. The organic feed contains some HNO3 which
gets stripped from the organic in the contactor thus increasing the acidity of the aqueous
phase and helps in preventing the hydrolysis of Pu. The interphase is generally
maintained at top (Aq. continuous - organic dispersed). The aqueous to organic flow ratio
is 2.5 : 2 .
Strip Acid
0.1 M HNO3
Loaded Solvent
Strip Section
Feed Organic
Loaded Organic U & Pu
Aqueous Product U & Pu
Fig. 4
The solvent after this cycle is generally sent for alkali washing to remove the
degradation products of TBP if any before recycling. The aqueous product is taken to
evaporators for concentration and after conditioning, they undergo a second cycle of
purification called partition cycle. Care should be taken to avoid any organic entrapment
26
in aqueous phase, which other wise will decompose in evaporators leaving solid residues.
These residues have a tendency to absorb and accumulate Pu.
PARTITION CYCLE
The Pu and U solution from intercycle evaporator is conditioned with NaNO2 to adjust
the Valency of Pu to tetravalent state and its acidity is again adjusted to 2-3M. The U and
Pu are once again extracted into organic as before and the loaded organic is subjected to
scrubbing in a separate column, if necessary. This loaded solvent is then taken to
partitioning contactor where U and Pu are separated from each other (which is the main
aim of this step). This is achieved in partitioning contactor by back extracting Pu from
TBP phase by reducing it trivalent state with an aqueous solution containing a suitable
reducing agent. Ferrous sulfamate or uranous nitrate stabilized with hydrazine are the
two most commonly used reducing agents - and these two methods will be discussed here.
(1) FERROUS SULFAMATE AS REDUCING AGENT (FE (NH2SO3)2:
Pu (IV) can be reduced by Fe(II) according to the following reaction.
Pu4+ + Fe2 <=====> Pu+3 + Fe3+
While Fe++ ion serves as reductant, sulfamate ion acts as a stabilizer for Fe(II) and
as a destroyer of nitrous acid. If the sulfamate ions were not present, small quantities of
nitrate ion always present in HNO3 would autocatalytically oxidize the ferrous ion thus
preventing Pu reduction. A several fold excess of ferrous sulfamate is used to ensure
complete reduction.
NH2SO3- plus NO2- <=====> N2H2O plus SO4Prevention of uranium extraction into aqueous phase is effected by increasing the aqueous
acidity in reductant above 0.5 M in HNO3.A very high aqueous acidity should be avoided
as Pu+3 and Fe++ are not stable at high acidities. Generally the loaded solvent is fed to the
bottom of the partition contactor and partitioning agent (0.03M Fe++ sulfamate 0.6M
HNO3) is fed from top. All the Pu is stripped as Pu is stripped as Pu+++ and leaves the
contactor at the bottom as aqueous product containing little Uranium.
The organic solvent containing U with little Pu leaves from top of the column and the
U from this solvent is stripped in a separate contactor using 0.01 M HNO3.
(2) URANOUS NITRATE AS REDUCTION AGENT:
Although ferrous sulfamate is used widely as reductant in PUREX PROCESS, it has
the disadvantage of introducing Fe and sulfate (produced by the hydrolysis of sulfamate)
into the solution thus causing an
1. increase in volume of waste to be stored.
2. When these solutions are evaporated the sulfate concentration also increases
causing corrosion problems.
3. Fission products like Sr and Ba from ppts with SO4.
4. Also sulfate complexes Plutonium very strongly making further purification of Pu
very difficult. Due to these reasons uranous nitrate stabilized with hydrazine has
27
been proposed as an alternate reductant for the partitioning of U and Pu and
production flow sheets have been developed for its use. Various parameters
governing the effective use of uranous are given below.
Uranous nitrate alone if used for the reduction of Pu (IV) is highly unstable. HNO2
always present in extraction contactors oxidized Pu (III) and U(IV) . The oxidation of
Pu(III) by NO2-is faster than the oxidation of U(IV) to U(VI) by NO- Pu oxidation by
HNO2 is almost instantaneous . In the absence of a stabilize the oxidation of U (IV) by
NO2-is auto catalytic as can seen from the following reactions.
U+4 + NO3- + H2O <=====> UO2++ + HNO2 + H+ Slow
U+4 + 2HNO2 <====> UO2++ + 2NO + 2H+
Fast
2NO + NO3- + H2O <======> 2H+ + 3NO2Thus about three moles of NO- is produced for each mole that is consumed and the
reaction proceeds faster as concentration of NO2- increases.
Pu(III) reaction as follows:
Pu+3 + HNO2 + H+ <=====>Pu+4 + H2O + NO
Fast
Due to these reasons a stabilizer is always used along with uranous nitrate. A
substance that removes or destroys NO- is called a stabilizer it should react faster than
Pu(III) with HNO2. Sulfamate, Hydrazine (N2H4), Urea, Formaldehyde etc., have been
tried and hydrazine has been found to be very effective. HNO2 is always present in the
organic phase containing U and Pu and it has been found 0.1 M N2H4 removes 99% NO2from 30% TBP in 30 seconds.
N2H5+ + NO2 - <======> HN3 + 2H2O
HN3 + H+ + NO- <=======> N2O + N2 + H2O
HN3 is a hazardous chemical, its concentration remains very little if 0.1M hydrazine is
used as stabilizer. Acidity of the aqueous phase is also an important factor in the use of
U(IV) . The aqueous acidity should not fall below 0.5 M as the U(IV) is unstable below
that acidity. It should not also be above 2M as this would lead to the reoxidation of
Pu(III) increasing the possibility of Pu reflux.
Following reactions are involved.
U+4 + NO3- + H2O <====> UO22 + + HNO2 + H+
Slow
2Pu3 + + NO3- + 3H+ <======> 2Pu+4 + HNO2 + H2O
The concentration for uranous nitrate is also an important factor in the operation.
Uranous is more susceptible to side reactions leading to its oxidation which because of its
reasonable extractability in TBP, can also readily occur in organic phase. In fact uranous
has very limited stability in organic and is destroyed at a faster rate in the organic phase
28
than in the aqueous phase in contactors. Its extractability in the organic phase is an
important difference between uranium (IV) and Fe(II) (the latter is not extractable).
This means that a TBP scrub can remove the excess U(IV) being carried by aqueous
product Pu(III) stream and so reduce the contamination of the latter. This is one of the
major advantages claimed for U(IV). It is not only that we obtain purer Pu, we also obtain
a purer effluent from the subsequent Pu purification. U(IV) though extractable is less
extractable than could be wished. If we accept the limitation of the acidity to about 2.5
as to avoid the reoxidation of Pu(III), then the scrubbing of U(IV) of aqueous phase
requires bigger volume of TBP or more stages or both under these conditions U(VI)
is scrubbed out easily. The need to remove uranous tends to dictate the condition in the
TBP scrub section of the contactor.
Because of the extractability U(IV) it has to be introduced at the center of the
contactor at or near the feed point U , Pu stream. This brings it into immediate contact
with Pu and from here it is carried in both directions by the two streams giving reducing
agent in both phases of all the stages of contactor. If on the other hand it were to be
introduced with aqueous scrub like Fe(II) much of it would be extracted and removed
immediately, without reaching the plutonium rich stages.
Because of the various factors discussed above the quantity of uranous used is always
more than the stoichiometric amount required for the reduction of Pu(IV). Generally the
quantity of U(IV) used need not exceed four times that required stoichiometrically for
Pu reduction. But this ratio varies from contactor to contactor. Quite small quantities of
N2H4 (0.1M) in the aqueous scrub suffice to keep the system free from HNO2.
The general scheme used is given below. After the partitioning uranium bearing
organic stream is stripped of uranium in another contactor as discussed previously. The
aqueous plutonium product is sent for final purification.
TBP Scrub with U
Aq. Scrub 0.1 M N2H4
Aq. Scrub Section
O/A = 9/1
U(IV) Aq. Feed
Org. Feed TBP
U & Pu
TBP Scrub Section
TBP Scrub
Aq. Product Pu (III)
Fig. 5
29
THIRD CYCLE PURIFICATION OF URANIUM
The somewhat dilute uranium product from partitioning cycle is concentrated in
evaporators. The feed is adjusted to about 1M in HNO3 and ferrous sulfamate is added to
reduce Pu(IV) to Pu(III) and then it is fed to compound contactor where U is again
extracted by TBP to a higher saturation level. Pu(III), fission products remain in
aqueous phase. Scrub solution also contains ferrous sulfamate. The loaded solvent is
stripped in another column with 0.01M HNO3 and the aqueous product U is evaporated
and then passed on for tail end polishing, if required.
TAIL END PURIFICATION & CONCENTRATION OF PLUTONIUM USING TRILAURYL AMINE:
INTRODUCTION:
High molecular weight tertiary amines can react with anions (plus hydrogen ions) to
form salts that are relatively insoluble in aqueous solutions but soluble in organic solvents
such as benzene, xylene, chloroform and kerosene. These solutions of amines in a
hydrocarbon diluent extract acids from aqueous solutions to form alkyl ammonium salts
as shown in equation:
R3 N org. + HX aq. <=====> R3 NH x org
Acting as liquid anion exchange agent, these salts can readily exchange their anions
for another as shown
R3 NHX org + YAQ-
R3 NHY org + x Aq
The order of preference in the amine phase is actually similar to that of anion
exchange resins, namely, in decreasing order CLO4 > NO3 > Cl > HSO4 > F. These salts
behave like ion exchange resigns of weak base type and by contact with basic solutions,
the free amine is regenerated:
30
R3NHX org + OHAq-
(R3N) Org + H2 O + XAq-
The mechanism of ion absorption on this kind of exchanger is not fully clarified, either
simple exchange as in equation (a) or the adsorption of a neutral complex as in the
equation (b) is possible.
(Pu(NO3)6)- + 2(Haq+ ) + 2(R3 NHNO3)org <====>
(R3NH)2 Pu(NO3)6 org + 2Haq+ + 2NO3- aq
---
Pu(NO3)4aq + 2(R3N HNO3)org<=====>(R3NH)2 Pu(NO3)6 org
(a)
eq. (b)
The regeneration of alkyl ammonia salt generally be obtained by eluting the metal
complex with dilute acid. Since elements relevant to fuel reprocessing technology such as
U, Pu and Th show a tendency to form anionic complexes depending on the type and
concentration of the acid solution, good separation can be obtained by proper choice of
aqueous medium. Other variables liable to affect the extraction behaviour are to structure
of the alkyl chain whether linear or branched, the presence of any substituents in amine
formula and nature of the diluents.
FACTORS AFFECTING PROCESS APPLICATIONS AND CHOICE OF DILUENT:
As In the case of TBP, amine extractants are used in solutions of suitable
hydrocarbon diluents principally for reasons of density, viscosity and surface tension.
Other than general properties required (as discussed previously for TBP) a diluent must
show a reasonable ability to dissolve the amine metal complexes. The free base forms of
amines are readily soluble in a number of organic diluents but in many cases amine
salts with mineral acid show limited solubility. When solubility is exceeded the amine
salt precipitates yielding a solid or another organic phase. Factors influencing the
solubility of amine salts in the diluent are
(1) Type of amine and molecular structure
(2) Type of amine salt (sulphate > chloride > NO3)
(3) Polarity of diluent
(4) Acid / amine mole ratio in the organic phase
(5) Temperature
In general, the solubility of an amine salt in a diluent is better with a branched chain
amine than with straight chain compound: differences in structure are most important for
primary and secondary than for symmetrically tertiary amines. The degree of polarity of
the diluent strongly affects amine salt solubility, less polar solvents such as kerosene type
spirits and aliphatic and cycle parafinic compounds give lower solubilities than aromatic
type diluents. The addition of polar modifiers such as long chain alcohols to kerosene
type diluents improves the solubility limits. The tendency towards third phase formation
is decreased by an increase of temperature.
31
Moreover, the metal extractant complex resulting from long loading of the organic
phase with metal ions very often shown limited solubilities in the diluent. Again, the type
and structure of the amine and the polarity of the diluent determine this solubility. An
increase of diluent polarity is obtained by the adoption of aromatic diluents.
The effect of temperature on third phase formation in case of TLA nitrato complexes
of Pu(IV) is slight. As for the diluent is concerned, It can be controlled by a diluent
whose polarity is closer to that of the solute. Finally the influence of nature of diluent on
the extraction characteristics of the amine is considerable and in many cases important
variations in Ea* values are noticed when the same amine is used with diluents of
different nature.
To complete with the well-known tail end processes of plutonium, an amine
purification cycle must provide high decontamination from fission product and high
plutonium concentration. Also, the strip methods must avoid the presence in Pu effluents
of contaminants which are objectionable in the subsequent metal conversion step.
The feed is PUREX partition cycle Pu product stream containing 1 - 2 M HNO3 and
small quantities of SO4- and TBP, and Fe. Prior to amine extraction, removal of TBP and
oxidation of Pu(III) to Pu(IV) with NaNO2 are necessary.
With expected feed concentration of Pu at 0.5 - 2 g/L the derived product
concentration around 0.3M (20%). Suggested extractant is 0.3M TLA Shell Sol T.
Aqueous to organic ratios of 10:1 are possible with these system since Ea* for Pu(IV)
with a 0.3M TLA solution are very high (>>300) over a wide range of acidity, scrubbing
can be done with low acidity for better Ru DF.
Stripping of Pu from TLA is the main problem in amine based Pu purification
procedure. Stripping with acetic acid having 0.1M HNO3 seems to be very attractive over
various other methods available. The flow sheet proposed is given below which is still
under investigation.
Aq. Product
20-60 g/l
0.3 TLA in SST
E
X
T
N
Aq. Raff.
Feed 0.5-2 g/l Pu
1-2 M HNO3
2M HNO3
Scrub
S
C
R
U
B
2-4 M
HAC
0.1 M
HNO3
Org.
Product
S
T
R
I
P
P
I
N
G
Lean Org. for
Alkali Treatment
32
Fig. 6
FINAL CYCLE PURIFICATION AND CONCENTRATION OF PLUTONIUM USING
TBP
TBP processes have been proposed that concentrate and purify Pu by a technique
of total reflux with batch with drawl Pu that is stripped in the stripping contactor is
combined with the feed to the extraction contactor, after conditioning the valency state to
four. After a period of total reflux, Pu is with drawn in a batch by diverting the stripped
Pu. This technique is capable of a high degree of purification or decontamination as well
as concentration. By feeding back a part of the back washing product solution, it is
possible to obtain a Pu solution of constant concentration whatever the feed (Pu)
concentration for the purification cycle.
Using a pulsed column solutions containing 25 g/L Pu has been obtained. Uranous
nitrate stabilised with N2H4 has been used as stripping agent. The strip contactor alone
can give a concentration factor of 10.
The nitric acid concentrations and flow ratios chosen for the extraction, and scrubbing
of loaded solvent represent a compromise between conditions for concentrations of Pu
and for Zr - Nb decontamination. Ru decontamination is also adequate as the free
acidity is about 3 M in extraction section. A concentration factor of 3 can generally be
obtained in the extraction contactor. For further details please refer (11).
Pu XFM
3.5 M
Pu III
To Pu IV
100 V
LOADED ORG.
1M
HNO3
16.7 V
U IV 8M
H+ 1M
N2H4 0.1 M
HNO3 1M
N2H4 0.1M
2.29V
1.04 V
E
X
T
N
33.3V
30%
TBP
H
N
O
3
S
C
R
U
B
S
T
R
I
P
2.5V
Org.
33
Org
Scrub
30%
TBP
Raff.
PuO2 0.1M
H+ 0.1M
Aq.
Product
PURIFICATION AND CONCENTRATION OF PLUTONIUM
BY ANION EXCHANGE IN NITRIC ACID
INTRODUCTION
Few elements can match the diversity of ion exchange behaviour of Pu due to its high
ionic potential, Pu can be readily absorbed on to cation exchange resins. Also Pu is one of
very few elements which form anionic nitrate complexes and the absorption of Pu(IV)
out of nitrate solutions by anion exchange resin is highly specific for Pu and offers a
method for achieving a high degree of concentration and purification of Pu.
Dowex anion exchange resins are generally composed of polystyrene cross linked with
divinyl benzene. Permit SK is a copolymer, containing in addition to polystyrene, and
appreciable percentage of pyridinium groups. The exchange sites are normally strongly
basic quarternary alkylamnonium groups, in most cases, the trimethyl ammonium group N(CH3)3+. Exchange with these resins takes place by the following reactions.
R - N (CH3)3+ + Cl- + A- <======> R - N (CH3)3+ + A- + ClWhere A- represents any anion: permutit SK resins also contain quarternary
ammonium sites in addition to the moderately basic pyridinium. Physical factors have an
important influence on exchange properties. Since the rate of exchange is frequently
diffusion controlled, the use of small resin beds has a beneficial effect on the kinetics.
Exchange is also more rapid in resins of low cross linkages since they absorbs much more
water and thus allow diffusion to take place through an essentially aqueous rather than a
solid medium. The advantages of lower cross-links however are obtained at the expense
of greater swelling, lower selectivity and reduced resin stability. The percent crosslinking of Dowex resins is specified along with the name and number of the resin, thus
Dowex 1 x 4 refers to resin with 4% cross-linking.
Quantitative data on exchange equilibria are presented in terms of distribution co efficient which is defined as
Concentration of solute per gram of resin
34
Kd = -------------------------------------------------Concentration of solute per ml of solution
The anion exchange process for Pu purification consists of three steps.
(1) An absorption or loading stack in which Pu is absorbed out of a feed solution by
the resins.
(2) A washing step in which impurities are removed by washing the Pu laden resin
with an appropriate wash solution and
(3) An elution step, in which Pu is stripped off the resin in an appropriate aqueous
solution.
Absorption:
Plutonium in nitrate solution is apparently adsorbed on anion exchange resins as the
hexanitrato complex Pu(NO3)62- various factors affecting its absorption on the resin is
discussed below :
Acidity
Hexa nitrato species of Pu(IV) are formed in concentrated HNO3 solutions and
these are adsorbed on ion exchange resins. Fig given below indicates that the optimum
nitric acid concentration for the adsorbtion of Pu(IV) at 50 and 600 C is approximately
7M.
Decrease in the Kd value above 7M HNO3 is attributed to the formation of acidic species
according to the following equation:
Pu(NO3)6-2 + H+ <======> HPu(NO3)6HPU(NO3)6- + H+ <======> H2Pu(NO3)6
Since these species are less adsorbed their formation would lower the Kd in
concentrated acid. In Ca(NO3)2 however in spite of the decreased complexing,
adsorption is consistently higher than from HNO3 solution and there is no maximum in
the Kd curve and both of these effects are attributed to the absence of acidic complexes
mentioned above.
Temperature
Kd for the hexa nitrato species of Pu(IV) decreases with the increasing temperature
as shown in Fig. The capacity of resin also decreases to a certain extent. However, the
rate of absorption is faster at 600 C than at 250 C, which more than compensates for the
smaller Kd at higher temperature and it is generally recommended that the loading be
carried out at 500 C to 700 C.
Feed Pu concentration and loading:
35
The Pu concentration in 7 to 8M HNO3 feed was found to have little affect on resin
capacity above about 0.4 g Pu/L in the temperature range of about 25 to 600 C. Below 0.4
g Pu/L the equilibrium resin capacity decreases due to increased competition for the resin
site by NO3- ion.
Anion exchange resins "shrink" marked by up on absorbing Pu from HNO3.
Generally ion exchange resins swell when placed in water depending on the ionic nature
of the resin and strong ion pair formation will result in marked resin shrinkage. In case of
Pu, the absorption of large Pu(NO3)6 ion in place of two nitrate ion is responsible for
the loss of resin water and the resultant shrinkage. This marked loss of water from resin
during Pu absorption appears to be the cause of marked decrease in absorption rates with
increased resin loading.
EFFECT OF FEED URANIUM CONCENTRATION AND F.P.
U(IV) exhibits a Kd of 8 between 7M HNO3 and Dowex 1 x 4 (50-100 mesh) under
conditions which yield about a Pu - Kd of 3500. This means that U(IV) cannot compete
with Pu(IV0 for anion exchange resin under the conditions described unless the U(V)
concentration is high compared to Pu concentration. As the U to Pu ratio increases, the
resin capacity for Pu decreases because of increased competition of U. Until the U
concentration in the column feed reaches a value reaches several times greater than the Pu
concentration, no effect on column capacity is normally observed. Most of the fission
products are not absorbed.
WASHING OF THE LOADED RESIN:
The decontamination of Pu from fission products can be further improved by washing
the plutonium loaded resin with 7.2M HNO3 at 600 C but care must be exercised to avoid
loss of plutonium. The loss of plutonium during washing is affected by the total Pu
loading, and by the volume of the wash solution but not by washing rate. In about 20
column volume of washing almost all the U and F.P.s that have been weakly absorbed on
the column are washed out. At times, the presence of complexing agents like acetic acid
in very low concentrations in 7M HNO3 also help to increase the rate of removal of
impurities like Zr from the Pu laden resin during a washing cycle. Only those complexing
agents which would form a much stronger complex with fission products than with Pu are
useful.
ELUTION OF PLUTONIUM:
Four possible methods exist for the elution of Pu from an anion exchange column.
These are displacement by another anion, change of oxidation state, shift of equilibrium
by change in NO3- concentration and complexing of Pu(IV). Of these, least practical is
displacement by another anion, because of very restrictive requirement that the displacing
anion be very strongly held PU(IV) nitrate complex.
Elution by reduction with hydroxylamine, ascorbic acid and other reducing agents
have been tried. Generally elution by reduction is avoided because all the reductants tried
either caused severe gassing, introduced undesirable impurities precipitated the Pu or
were too slow to yield high product concentration.
36
A generally favoured elutriant is 0.5M (dilute) HNO3, because of the fact that no
impurities or expensive agents are added and no gassing occurs. Elution at 600 C gave
higher rate of elution and better product concentrations as compared to room temperature
operations. If the rate of elution with dilute HNO3 is fast, elution of a uniformly loaded
column will result in a very rapid rise in product plutonium concentration as the last of
the strong acid is replaced. The increase continues until the product solution concentration
reaches that limited by equilibrium of product solution with loaded resin. The Pu
concentration then remains constant at this equilibrium concentration until no more
loaded resin remains at the bottom of the column at which time it drops rapidly to 0. This
results in a square wave type elution curve. In practice, however when dilute HNO3 is
used as elutriant " Tailing " or continuous elution of Pu in small quantities in successive
column volumes have been found to occur often. This decreases the product concentration
and increases its volume (Further, the elution is carried out by upward movement of
elutriant to have richer product concentration)
Dilute acetic acid as an elutriant for Pu has been in use at Trombay for a long time. It
has a marked advantage over dilute HNO3 as most of Pu is eluted in 2 to 3 column
volumes no tailing has been observed. Further acetic acid does not interfere with further
processing of Pu like oxalate precipitation. 1M acetic acid 0.6M HNO3 is the normal
composition of the elutriant used. Pu is complexed by acetic acid and is removed rapidly
from the column. Most of Pu is eluted out in 2nd and 3rd column volume thus giving a
very high product concentration. This Pu solution is then sent for oxallate precipitation.
37
38
SUGGESTED BOOKS AND REPORTS FOR STUDY
Sr
Book Title
1
REACTOR HAND BOOK -VOLUME 2 -FUEL REPROCESSING
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
CHEMISTRY OF PLUTONIUM
CHEMICAL PROCESSING OF REACTOR FUELS
CHEMICAL PROCESSING OF NUCLEAR FUELS
CHEMISTRY OF ACTINIDE ELEMENTS
PROCESS CHEMISTRY VOL . II
PROCESS CHEMISTRY - VOL.4 SERIES III
PLUTONIUM
AQUEOUS REPROCESSING CHEMISTRY FOR IRRADIATED
FUELS
INT. CONFERENCE ON SOLVENT EXTRACTION CHEMISTRY
OF METALS -SEPT 1965
ENGINEERING OF FUEL REPROCESSING
SOLVENT EXTRACTION CHEMISTRY SYMPOSIUM
PLUTONIUM HAND BOOK VOL.I & II
SOLVENT EXTRACTION IN ANALYTICAL CHEMISTRY
ION EXCHANGE
A.E.R.E. - R - 4381 - U(IV) AS REDUCTANT IN U / PU
SEPARATION
DP. 700
BIBLIOGRAPHY ON SOLVENT EXTRACTION OF PLUTONIUM
INTERNATIONAL CONF. ON SOLVENT EXTRACTION
RECOVERY PURIFICATION AND CONCENTRATION OF
PLUTONIUM, BY ANION EXCHANGE IN HNO3
Author
EDITED BY S.M. STOLLER AND RICHARDS R.B. –1961
INTER SCIENCE PUBLISHERS
CLEVE LAND
EDITED BY JHON F. FLAGG
MARTIN & MILES
KARTY & SEABORG
BRUCE F.R.
EDITED BY STEVENSON & MASON
TAUBE
B. RUSSELS SYMPOSIUM
EDITED BY MACKAY & OTHERS
J.LONG
EDITED BY DRYSSEN , RYDBERG & OTHERS
EDITED BY O.J.WICK
FREISER & MORRISSON
BY HELFERICH
H.A.C. McKAY AND OTHERS
L.L.SMITH
(HELD AT HAGUE , 1970)
J.L.RYAN & E.J.WHEEL, WRIGHT H.W. 55893 -JAN.1959
39