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Патент USA US3046097

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July 24, 1962
F. R. BRUCE
3,046,087
SOLVENT EXTRACTION PROCESS FOR SEPARATING URANIUM
AND PLUTONIUM FROM AQUEOUS ACIDIC SOLUTIONS
OF NEUTRON IRRADIATED URANIUM
Filed Jan. 15, 1956
5 Sheets-Sheet 1
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3,46,687
Patented July 24, 1962
2
traction and maintain a single, reproducable distribution
co-e?icient between organic and aqueous phases. Thus,
decontamination with respect to ruthenium has proven
to be the limiting factor in the decontamination of
neutron-irradiated ?ssionable material. For example, ap
3,046,087
§0LVENT EXTRACTION PRGQESS FOR SEPA
RATING URANIUM AND PLUTQNIUM FROM
AQUEOUS ACIDIC SQLUTIONS 0F NEUTRON
IRRADIATED URANIUM
Francis R. Bruce, Oak Ridge, Tenn, assignor to the
United States of America as represented by the United
proximately 75~90% of the remaining ‘beta activity after
States Atomic Energy Commission
Filed Jan. 13, 1956, Scr. No. 559,080
6 Claims. (Cl. 23-145)
was due to ruthenium.
the ?rst cycle of a prior art solvent extraction process
To improve ruthenium decontamination, special pre
10 treatments (prior to solvent extraction processing) have
My invention relates to an improved process for the
decontamination of a neutron-irradiated ?ssionable and
fertile material and more particularly to an improved
solvent extraction process for such decontamination.
In utilizing uranium and plutonium as fuels in nuclear 15
been devised. In one method, described in the co-pend
ing application of the common assignee S. N. 561,962
?led January 27, 1956, in the name of Allan T. Gresky
reactors, they would ideally be left in the reactor until
substantially all the ?ssionable material has been con
sumed by ?ssion. In practice, however, the fuel is with
drawn from the reactor for decontamination from ?ssion
products long before it has been totally consumed. For 20
example, uranium having the natural isotopic concentra
the solution is subjected to treatment with acetone and
tion may be withdrawn from a reactor after the concen~
tration of uranium-235 has been reduced from an initial
0.71% to only approximately 0.64%. This is done to
prevent the accumulation of excessive quantities of ?s
sion products having large neutron absorption cross
sections. An extremely small amount of such ?ssion
products has a highly deleterious effect on the reactivity
of the reactor and may even threaten the continuance of
the chain reaction.
for “Improved Ruthenium Decontamination Method,”
now US. Patent No. 2,945,740 issued July 19, 1960,
sodium nitrite prior to processing.
However, such
methods are time consuming and burdensome, and since
they are essentially unit operations, they slow down and
impede continuous chemical processing of reactor fuels.
The term “?ssionable material” as used herein refers
to uranium-235, uranium-233 and plutonium, and the
term “fertile material” refers to thorium and uranium
238. The term “?ssion” refers to the splitting of uranium
and plutonium into a plurality of parts upon the capture
of a neutron of appropriate energy, and the term “?ssion
products” refers to the immediate product nuclei from
?ssion as well as to their radioactive decay products.
(See Glasstone, op. cit., espectially pages 105-128.) The
Furthermore, when the reactor is 30 closely similar statistical ?ssion product yields of U—233,
U—235 and Pu-239 are shown in Stevenson, Introduction
to Nuclear Engineering, page 50. Glasstone, op. cit.,
pages 389—397, indicates the ?ssion product species of
major importance in fuel reprocessing after different
1y rapid rate relative to the production thereof, with a 35 lengths of radioactive decay.
resulting decrease in yield. Since the ?ssionable material
In view of the di?iculties experienced by the prior art
remaining in the spent fuel element constitutes a signi?
in solvent extraction decontamination of ?ssionable ma
cant and valuable quantity that may be re-used directly
terial, particularly in decontamination with regard to
employed to produce uranium-233 or plutonium as a
primary product, the new ?ssionable species must be re
moved before they are permitted to concentrate to a
point at which they undergo ?ssion at an uneconomical
as a reactor fuel, reactor design permitting, or further
ruthenium, it is an object of my invention to provide an
concentrated by such isotopic separation means as gase 40 improved method for the recovery and decontamination
ous diffusion, the economical recovery and decontami
of neutron-irradiated ?ssionable and fertile material.
nation of such fuel is of supreme importance in the
Another object is to provide an improved method for
development of an atomic energy program.
the decontamination of ?ssionable and fertile material by
The processing of reactor fuel differs from most
solvent extraction.
chemical processing principally in that minor quantities 45 Another object is to provide such an improved method
of ?ssion products must be separated from large quanti
ties of substantially unchanged material. The chemical
processing associated with the operation of nuclear re
actors, therefore, generally has three primary objectives:
in which decontamination from ?ssion products, particu
larly ruthenium, is greatly improved, while yet not dimin
ishing ?ssionable material recovery.
Still another object is to provide an improved contin
removal of fission product poisons from the remaining 50 uous solvent extraction process for such fuel recovery
fuel; the reclamation of the fuel; and the recovery of
and decontamination, wherein decontamination with re
uranium-233 or plutonium when desired.
gard to ruthenium is vastly improved without an inter
For general information concerning the processing of
posed special unit operation.
nuclear reactor fuel, reference is made to Glasstone,
Yet another object is to provide such a process which
Principles of Nuclear Reactor Engineering, especially to 55 displays a high degree of mechanical operability and
Chapter 7 and pages 416-428.
Prior art solvent extraction processes, commonly con
ducted in aqueous nitric acid systems, employed strongly
acidic conditions, say in the order of two normal nitric
acid. While uranium recoveries were satisfactory, de~ 60
contamination from ?ssion products generally, and par
ticularly ruthenium, left much to vbe desired. As Glass
?exibility and which is capable ‘of remote control oper
ation.
These and other objects and advantages of my inven
tion will become apparent from the following detailed
description, the accompanying drawings and the attached
claims.
In accordance with my present invention I have pro
vided, in a solvent extraction process for decontaminating
tone states on page 296 of the cited work, ruthenium is
easily considered the most dii?cult element to separate 65 neutron-irradiated ?ssionable and fertile metals which in
cludes the selective extraction of at least one ?ssionable
from the desired products in fuel processing. The reason
is that, 'in addition to having amphoteric properties, it
exhibits several, possibly six, oxidation states. In view
or fertile metal from an aqueous mineral acid solution
with a substantially water-immiscible organic solvent, the
improvement which comprises conducting said extraction
of this, and the tendency of ruthenium to exist in various
forms of molecular association, such as in colloids and 70 under conditions‘ of a net de?ciency of total acid-forming
anions, i.e., a slight stoichiometric de?ciency of inorganic
polymers, it has heretofore been extremely dif?cult to
anions other than hydroxyl. Since solvent extraction
con?ne ruthenium to a single phase during solvent ex
processes are generally conducted in nitric acid solutions,
8,046,087
3
this amounts to a nitrate ion de?ciency. For convenience
in presentation, my invention will, therefore, hereinafter
be illustrated speci?cally with regard to nitrate ion de?
cient solvent extraction conditions.
The employment of my invention vastly improves
decontamination with regard to ?ssion products in solvent
extraction processing of reactor fuel. For example,
an example, consider an aluminum nitrate solution that
is nitrate ion de?cient. Stoichiometric amounts of alu
minum and nitric acid are contacted to form aluminum
nitrate, or aluminum nitrate itself is employed. When
additional amounts of aluminum are added to such a
solution the water reacts with aluminum ions to form
aluminum hydroxide.
The aluminum hydroxide may
then exchange ions with the aluminum nitrate present,
ruthenium decontamination factors under my nitrate ion
resulting in the formation of aluminum hydroxy or di
de?cient conditions are commonly 103 as compared to
with ‘only 5 or 6 using acid ?owsheets. Total ?ssionable 10 hydroxy nitrates [Al(OH)(NO3)2 and Al(OH)2NO3].
If too much aluminum is added, insoluble Al(OH)3 is
material beta decontamination is commonly 6x105 as
formed. Similarly, hydroxy uranyl nitrate salts may be
compared with 2X 103 with an acid ?owsheet, and total
formed. The hydrolysis of these weakly acidic salts will
gamma decontmiination is similarly spectacular-4X 105
then supply hydrogen ions to the solution which gives
as compared with 1x103. My conditions surprisingly
do not effect plutonium and uranium recovery, contrary 15 the solution an acidic pH. ‘(Of course, the same hy
drolysis mechanism applies if preformed basic aluminum
to what might be expected; plutonium and uranium re
or uranium nitrate salts are used.) Thus, in nitrate ion
covery of 99.9% is generally obtainable. No special
de?cient systems, the hydrogen ions are supplied to the
unit-operation preceding solvent extraction is necessary,
solution from the hydrolysis of weakly acidic salts rather
and fully continuous chemical processing is practical. My
nitrate ion de?cient extraction conditions are not limited 20 than directly from the introduction of nitric acid. Gen
in applicability to any single ?owsheet or combination of
?owsheets or to_ the recovery of any single isotope of
uranium or plutonium. Furthermore, in processes for
recovering the neutron-irradiated fertile material, thori
um, my method likewise aids in its recovery together
with that of the ?ssionable material which is bred from
it, uranium-233. My method has been very successfully
employed as a principal feature in a number of current
large scale production processes for the recovery of ?s
sionable and fertile material.
Having been so success
fully demonstrated to date, it shows further promise of
improving reactor fuel recovery, thereby directly con
tributing to the advancement of economical nuclear power.
As understood in this speci?cation and in the appended
erally, I ?nd that an aqueous nitrate ion solution of ?s
sionable (or fertile) material of approximately 0.05-0.5
normal nitrate ion de?ciency is satisfactory, while about
0.3 normal nitrate ion de?ciency is preferred. It is ap
preciated that with extreme nitrate ion removal, hydrolysis
effects ‘will continue until the solution becomes basic and
uranium, aluminum and other precipitations occur.
In FIG. 1 is shown a nomogram of an aqueous uranyl
nitrate solution, which correlates uranium molarity and
nitrate ion de?ciency to pH. On this scale negative values
refer to nitrate ion de?ciency, the 0 point represents
stoichiometric equivalency, and the positive values indi
cate the amount of free acid. To ?nd the third parameter
when two are given, draw a line connecting the two
claims, nitrate ion de?ciency is a relative term to denote 35 given points and extend it, if necessary, to give the third
reading.
the condition of a solution in which more equivalents of
In addition to formulating nomograms such as in FIG.
total metal ions, usually weakly basic, are present than
1, the nitrate ion de?ciency of my reactor fuel solutions
equivalents of nitrate ions. Thus, such a solution of a
may be analytically determined. Although the exact
nitrate salt of a given molarity will not register as high
analytical technique employed is not critical, the follow
an acidity as a solution of the normal nitrate salt of the
ing is one suitable method. It involves titration with
same metal molarity, or in other words, this is a measure
standardized alkali, after complexing polyvalent metal
of a stoichiometric de?ciency of nitrate ions. This stoi
ions with oxalate. The reagents are saturated potassium
chiometric de?ciency is cured through addition of hy
oxalate solution, 0.1 N NaOH standardized against po
droxyl ion by hydrolysis or direct base addition, rather
than by direct addition of other anions such as sulfate; 45 tassium acid phthalate and 0.1 N HCl standardized against
the foregoing NaOH. An aliquot of sample is pipetted
in this regard nitrate ion de?ciency should be understood
into a titration vessel and a small magnetic stirring bar
to re?ect the total anion (other than hydrox‘yl) de?ciency
is placed in the vessel. If less than 5 ml. of a 0.1 N NaOH
of the system. Hence, when it is said that a solution of
solution would be required to neutralize the estimated
Weakly basic or amphoteric cations is 0.1 normal nitrate
ion-de?cient, it is understood that the solution contains 50 acidity of the sample, pipette an HCl spike into the titra
tion vessel. Next, pipette 10 ml. of the potassium oxalate
that much less nitrate ion than the solution of the norm-a1
solution into the vessel, buffer a Beckman automatic titra
nitrate salt of the same metal, regardless of the metal
tor, set the pH dial to read 7.0 and titrate with the NaOH.
molarity of the solution; that is, a one molar uranyl
The calculation to give the total milli equivalents of ni
nitrate solution and a three molar uranyl nitrate solution
may each ‘be 0.1 normal nitrate ion de?cient, since both 55 trate ion de?ciency in the sample is:
solutions lack just that amount of nitrate ion to meet
stoichiometric requirements.
However, the acidities of
(ml. of baseXN of base)—(ml. of spikeXN of spike)
It should be noted here that in extraction both feed
and
scrub solutions need not be nitrate ion de?cient,
salt concentrations.
Nitrate ion de?cient uranium, thorium and aluminum 60 provided the net extraction conditions are nitrate ion de
?cient. Thus, either the feed or scrub solution may be
solutions (aluminum nitrate is commonly used as a salt
acid, as long as the other solution is sufficiently nitrate
ing agent in solvent extraction, as will be shown below),
ion de?cient to overcome it and give the required net ni
may be conveniently achieved by dissolving additional
trate ion de?ciency. Generally, however, both nitrate
uranium, thorium or aluminum metal in aqueous solu
tions of the normal salt, by boiling off nitric acid as nitro 65 ion de?cient feed and scrub conditions are preferred.
The choice of the organic solvent for extracting the
gen oxides, or in the case of aluminum, by directly em
the solutions will not be the same in view of the different
ploying a basic salt such as aluminum hydroxy or di
hydroxy nitrate. This result may also be brought about
by direct addition of ‘a base, such as sodium or preferably
ammonium hydroxide.
Although the chemistry of nitrate ion de?cient solu
tion is not completely known, and I do not wish to be
bound to any particular theory, an attempt will be made
to explain how nitrate ion de?ciency is obtained and its
effects on decontamination from ?ssion products. As 75
fertile \and/ or fissionable material from my nitrate ion
de?cient aqueous feed solutions, while suppressing ?ssion
products extraction, is subject to considerable varia
tion. The choice depends upon the selectivity of the
solvent for the desired products, the ease with which
products can be stripped from the organic material after
the primary extraction, chemical and radiation stability
of the solvent, immiscibility of the organic-aqueous mix
ture, speci?c gravity of the organic as compared with the
amigos"?
5.
6
aqueous solution and viscosity. Among the satisfactory
solvents are ethyl ether, penta ether, and diiosopropyl
carbinol. For most ‘purposes, however, hexone (methyl
and in addition is not extracted as well as Pu (VI); ex
traction of plutonium is therefore preceded by oxidation
of plutonium to the VI state with alkali dichromate at
elevated temperature. The feed solution is adjusted to
isobutyl ketone) and tri-n-butyl phosphate (hereinafter
called “TBP”) are particularly satisfactory. Hexone may
be used without any diluent, while TBP should be diluted
with an inert, saturated hydrocarbon diluent, preferably
approximately 1-4 molar uranyl nitrate and 0.1-0.5 nor
mal nitrate ion de?ciency, and is made approximately
0025-05 molar in dichromate to oxidize the plutonium.
a kerosene fraction. Para?inic kerosene fractions are pre
In the extraction column, ‘both uranium and plutonium
ferred.
are extracted with hexone ‘While con?ning ?ssion products
The contacting of the nitrate ion-de?cient solution of 10 to the aqueous phase, and the organic phase is scrubbed
neutron irradiated ?ssionable and fertile material with the
in the same column with an aqueous aluminum nitrate
organic extractant may be performed in various manner.
solution of approximately the same nitrate ion de?ciency
For example, it may be performed batchwise in separa
as the feed solution and containing a slight amount of
tory funnels or in mixer-settlers. In column operation,
dichromate, less than approximately 0.1 molar. In the
packed, perforated-plate, pulse columns or the like may 15 second column the plutonium is stripped from the organic
‘be used. Continuous column operation is naturally pre
phase with approximately 0.01—0.2 molar ferrous sulfa
ferred for large scale operation. Since countercurrent
m-ate, which reduces plutonium to the inextractable III
operation provides. for most ef?cient mixing, the aqueous,
state without effecting the uranium. To prevent uranium
nitrate ion de?cient feed solution is introduced about the
stripping, the aqueous stripping solution is made approxi~
middle of the column, and the organic extractant is in 20 mately 0.5-3.0 molar in aluminum nitrate. This aqueous
troduced at the bottom (assuming the organic phase has
~solution is scrubbed with additional, slightly acidi?ed
a speci?c gravity less than one; otherwise the points of
hexone ‘before being withdrawn from the bottom of the
introduction are inverted). ‘At the top of the column,
column. The uranium is then stripped from the organic
Where the organic phase is withdrawn, it is highly bene
phase in the third column with slightly ‘acidi?ed water,
?cial to scrub any extracted ?ssion productsfrom the 25 say 0.01—0.15 molar in nitric acid.
organic phase with an aqueous scrub solution of a nitrate
At higher radiation levels, it may be desired to put the
ion de?cient scrub solution, of about the same nitrate ion
uranium product stream, and possibly the plutonium prod
de?ciency as the feed solution. While various inorganic
uct stream through several additional solvent extraction
nitrate salts may 'be used, aluminum nitrate is especially
cycles. The uranium product stream from the ?rst cycle,
noteworthy, because of its e?icient salting action, the
ease in which nitrate ion de?cient solutions may be ob
tained, and because aluminum, which is a common jacket
30 which may be less than one molar in uranium, is evapo
rated to approximately l.5—3.5 molar uranyl nitrate and
0.1-0.5 normal nitrate ion de?ciency, and is put through
ing material for uranium reactor fuel, may already be
a second uranium cycle which is nearly identical with the
conveniently present in the feed solution. If natural
?rst cycle except that plutonium is not signi?cantly pres
uranium or only slightly enriched uranium is being proc 35 ent and the uranium is stripped from the organic extract
essed, such that considerable plutonium may be present
in the second column and the aqueous scrub solution is
and its recovery desirable, this may ‘be accomplished by
approximately 0.02-02 molar in the reductant, ferrous
oxidizing the plutonium to the hexavalent stage to pro
sulfamate, and has no oxidant, in order to purify the
mote its extraction by the organic phase. Then, in a
uranium with regard any trace amounts of plutonium.
second column the uranium and plutonium may be sepa 40
If additional plutonium decontamination is required,
rated in the organic extract by preferentially reducing
the plutonium product stream from the ?rst cycle, already
the plutonium to the tri-positive state,-such that it may be
salted with aluminum nitrate, is oxidized with dichromate
stripped from the organic phase with an aqueous strip
and then decontamined ‘by at least a second solvent ex
ping solution. The uranium, which then remains in the or
traction cycle similar to the ?rst except that no signi?
ganic phase, may be introduced into a third column where 45 cant amount of uranium is present. The plutonium is ex
it may be stripped with an aqueous solution under the
tracted with slightly acidi?ed hexone, is scrubbed with
proper conditions. In the event that plutonium recovery
is not desired or if highly enriched uranium-235 is being
a nitrate ion de?cient aluminum nitrate solution provided
with dichromate, and then is stripped with dilute nitric
processed, such that little plutonium is present, then the
acid in a second column. Ferrous sulfamate is not re
uranium may be simply stripped from the original organic 50 quired in the scrub stream, as in the ?rst cycle partition
extract in a two-column operation. If higher decontami
ing column, where separation from uranium is accom
nation rates are required, separate second or third-cycle
plished. The plutonium solution from the strip column
plutonium and/ or uranium extraction cycles may be per-'
formed in the manner of the ?rst cycle extraction and
stripping columns.
of the second cycle may then be salted with aluminum
nitrate and put through a third cycle, similar to the
55 second.
With this “background of column operation, ‘a general
description will now be given of three diiferent, speci?c,
solvent extraction processes particularly developed about
The process losses in this process for plutonium are
less than 0.2%; uranium losses are less than 0.1%. The
plutonium content of the uranium stream is approximately
one part in 108 parts of uranium after two cycles of
my nitrate ion de?cient feed conditions. The ?rst is a
process for the recovery and decontamination of uranium 60 extraction.
and plutonium from‘ natural or slightly enriched uranium.
Decontamination from ?ssion products and
ruthenium is extremely high, in the order of 105.
The second is a process for the recovery of highly en
riched uranium. The third is a process for the separation
of protactiriium, thorium and uranium-233 ‘from neutron—
The second solvent extraction process built upon my
nitrate ion de?cient feed conditions is for the recovery
of uranium highly enriched in uranium-235. This process
irradiated thorium. More detailed description of these 65 is similar to the previous process except that separation
processes will ‘be found in the speci?c examples, but the
and decontamination of plutonium is not required since
process outlines will be given here in order to illustrate
the versatility of nitrate ion de?cient extraction conditions
in processing varying types of reactor fuels and fertile
materials. In the ?rst process, both uranium and plu 70
tonium are extracted by hexone from aqueous nitrate solu
only small amounts of plutonium are produced in en
riched uranium fuels. Any small amounts of plutonium
in the feed stream follows ?ssion products into the aque
ous Waste from the extraction column. Any aluminum
present as a uranium diluent and cladding material in the
fuel element diminishes the fresh aluminum requirement
in the scrub solution and serves as av salting agent in the
tion, while ?ssion products are only very slightly extracted.
At the nitrateion de?cient condition required for best
decontamination and also for solvent stability, Pu (IV)
extraction step. The process essentially comprises the
may hydrolize to the non-extractable polymeric ‘form, 75 following steps: dissolution of uranium or uranium~alu
3,046,087
8
minum alloy in 60% nitric acid with mercuric nitrate
catalyst for aluminum dissolution (approximately 1%, by
weight, of the aluminum); feed clari?cation by ?ltration;
feed adjustment to approximately 0.5-3.0 molar alumi
cated by the heavy line. PEG. 3 shows the ?rst cycle for
the separation and decontamination of uranium and plu
tonium and FIG. 4 shows a second uranium cycle. A
second plutonium cycle was conducted with the {BF
num nitrate, 0.05-0.5 normal nitrate ion de?ciency; and
separation of the uranium from ?ssion products and any
plutonium in at least one cycle of solvent extraction, using
present.
stream as in the FIG. 3, except that uranium was not
Plutonium activity in the uranium product was reduced
to ‘approximately one part in 108 parts of uranium after
hexone as the solvent. Any trace amounts of plutonium
two cycles. The uranium recovery was over 99.9%.
are separated from the uranium in a second cycle after
Table 1, below, shows the decontamination factors for
being reduced to the inextractable trivalent state with fer 10
uranium and plutonium achieved in two uranium and
rous sulfamate which is added with the aqueous, nitrate
plutonium cycles of solvent extraction.
‘
ion de?cient, aluminum scrub solution. The small quan
tities of plutonium are discarded with the ?ssion products.
Table I
Excellent uranium decontamination is achieved, with
DECONTAMINATION FACTORS
15
99.9% recovery.
The third process developed about my nitrate ion de
log D.F.
?cient solution is for the separation of protactinium,
thorium and uranium. from neutron irradiated thorium,
Constituent
1st Cycle
2nd Cycle
U
U
such as may be used in a breeder program for converting
For details 20
fertile thorium to ?ssionable uranium-233.
concerning this process, reference is made to the co-pend
ing application of the common assignee, S.N. 602,686,
?led August 7, 1956, in the names of A. T. Gresky et al.,
for “Process for Separation of Protactinium, Thorium and
Uranium from Neutron-Irradiated Thorium.” Brie?y, in
Gross alpha ________________________ _.
this process an aqueous thorium nitrate solution of neu
tron-irradiated thorium is adjusted to feed conditions of
approximately: 0.5-3.0 molar thorium nitrate, '0.25-1.5
molar aluminum nitrate and (Ll-0.6 normal nitrate ion
de?ciency.
The aqueous feed is introduced near the
middle of the extraction column, the extractant, approxi
mately 42% TBP—58% Amsco 125-82 (an inert paraf
?nic, kerosene-type diluent), ?ows upwardly through the
3. 7
Pu
3. 9
3.9
4.3
Y 7.0
6.5
1 8
1 8
Pu
2. 9
2.9
______________ __
4. 3
4. 0
6. 3
...... __
2. 7
4. 3
3. 3
3. 8
4. 7
6. 3
2. 8
2.8
When the above process was run under acid conditions
the ruthenium decontamination factor was several orders
of magnitude less, and the total beta and total gamma
decontamination factors were each two orders of magni
tude less.
,
column and extracts the thorium and uranium-233. An
EXAMPLE II
This example is intended to show a large scale produc
aqueous solution of approximate composition 0.2-1.5
tion run from the recovery of uranium from highly en
molar aluminum nitrate, 0.1-0.6 normal nitrate ion de
?ciency, 0.005-0.5 molar ferrous sulfate, and 0.00 1-0.010
molar phosphoric acid is introduced at the top of the
column to scrub any extracted protactinium and ?ssion
riched, neutron irradiated uranium. No plutonium re
covery was attempted, in view of the trace quantities
present in the enriched material. The procedure outlined
in the ?owsheets in FIGS. 5 and 6 was exactly followed.
40
products from the organic extract.
The uranium recovery exceeded 99.9%. Table 11 below
The aqueous phase from the extraction column, which
shows the decontamination factors achieved.
is about 0.1-0.6 normal nitrate ion de?cient, and contains
Table II
the protactinium-233, ?ssion products and other impuri
DECONTAMINATION FACTORS
ties is reduced in volume by evaporation to permit mini
mum storage volumes and/ or the high aluminum nitrate
log D. F.
concentrations that may be required for salting in subse
Constituent
quent protactinium-233 recovery cycles or for solvent
1st Cycle 2nd Cycle Over-all
extraction of the uranium-233 daughter.
The organic extract from the extraction column, con
Gross alpha ___________________________ __
4.04
2. 41
6. 45
taining thorium and uranium-233 is cascaded to the middle
Gross beta."
3. 84
2. 36
6. 20
of a second, partitioning column.- The thorium is stripped
Zr ...... ._
4. 43
3.04
7. 47
Nb. __
4.04
3. 58
7.62
into an aqueous phase of approximately 0.05—0.5 normal
Ce.___
4. 60
3. 99
8.59
nitric acid, and this aqueous solution is scrubbed by an
Ru. _ _
2. 55
2. 49
5.04
Pu
___________________________________
__
1.
30
1.
95
3. 25
organic stream of approximately 42% TBP-—5 8% diluent
fraction introduced at the bottom of the column. The 55
organic e?iuent from the partitioning column, containing
EXAMPLE III
all the uranium-233 and having a nitric acid concentra
For
examples
of
a
process for the separation of pro
tion of less than approximately, 0.01 normal is passed to
tactinium, thorium and uranium from neutron irradiated
a third column where the uranium is stripped into very
slightly acidi?ed water. The uranium may be concen 60 uranium, reference is made to the examples in the previ
ously identi?ed co-pending application of the common
trated and further decontaminated from the aqueous
assignee of Gresky et al.
product stream by an additional solvent extraction cycle
Although the above are examples of speci?c nitrate ion
or passage onto an organic cation exchange resin char
de?cient solvent extraction processes, the tremendous im
acterized by a plurality of nuclear sulfonic acid groups.
provement in decontamination factors in changing to my
The adsorbed uranium can be eluted from this resin with
nitrate ion de?cient conditions is strikingly shown over a
aqueous eluant.
range of values in the nomogram in FIG. 2. It can be
The following examples are offered to illustrate the
seen that orders of magnitude are involved. The same
foregoing processes in more detail.
values hold for all uranium concentrations, and similar
EXAMPLE I
70 improvement in beta decontamination, not shown, have
been obtained.
This example is intended to show an actual production
The above examples are only illustrative and should
scale process for the separation and decontamination of
not be construed as limiting the scope of my invention.
uranium and plutonium from neutron-irradiated uranium.
It can be seen from these examples, however, that my
The procedure outlined on the ?owsheets in FIGS. 3 and
4 was exactly followed. The main process ?ow is incli 75 invention is of great versatility and is inherently of wide
—
aoaaosr
applicability. Therefore, my invention is understood’ to
be limited only as is indicated by the appended claims.
Having thus described my invention, I claim:
1. An improved process for separating uranium and
plutonium from an acidic aqueous solution of neutron
irradiated uranium containing same together with ?ssion
products and nitrate ions, which comprises securing plu
tonium in its hexapositive state, contacting the resulting
it)
mately 1:8 molar aluminum nitrate and approximately
“0.2 normal nitrate ion de?cient; said aqueous plutoniun
stripping solution is approximately 0.05 molar ferrou:
sulfamate and 1.5 molar aluminum nitrate and sait
hexone scrubbing reagent is approximately 0.05 normal
in nitric acid; and wherein said aqueous uranium stripping
reagent is approximately 0.04 normal in nitric acid.
5. A process for recovering uranium from an acidit
feed solution, under net nitrate ion de?cient conditions,
aqueous solution of neutron irradiated uranium contain‘
with an inert, substantially Water-immiscible organic sol 10 ing same together with ?ssion products and nitrate ions
vent, scrubbing the resulting uranium and. plutonium
which comprises adjusting said solution to approximatel;
containing organic phase with an aqueous solution of
0.5-4 grams per liter of uranium, 0.5-3.0 molar aluminurr
aluminum nitrate, separating the scrubbed organic phase
nitrate, and 0.-05-0.5 normal nitrate ion de?ciency; con
from the resulting ?ssion products-containing aqueous
tacting the resulting feed solution with hexone, scrubbing
phase, contacting the separated organic phase with an 15 the resulting uranium-containing organic phase with at
aqueous nitric acid solution containing a plutonium re
aqueous solution approximately 0.5-3.0 molar in alumi‘
ductant, separating the resulting reduced plutonium
containing aqueous phase from the resulting uranium
num nitrate and 0.05-0.5 normal nitrate ion de?cient
separating the scrubbed uranium-containing organic phase
from the resulting ?ssion products-containing aqueous
2. The method of claim 1 wherein said feed solution 20 phase, contacting the separated organic phase with at
containing organic phase.
and the aqueous aluminum nitrate scrub solution are each
aqueous nitric solution less than approximately one nor‘
approximately 0.05-0.5 normal nitrate ion de?cient.
3. An improved process for recovering uranium and
mal in nitric acid, thereby stripping said uranium intc
products and nitrate ions, which comprises adjusting said
solution to approximately ‘1.0-4.0 molar uranyl nitrate,
cycle, comprising adjusting said stream to approximateh
the resulting aqueous uranium product stream.
plutonium from an acidic aqueous solution of neutron
6. The method of claim 5 wherein said uranium prod
irradiated uranium containing same together with ?ssion 25 uct stream is subjected to a second solvent extractior
150-300 grams uranium per liter and 2-3 normal nitric
0.025-03 molar dichromate ion, and 0.05—0.5 normal
acid, contacting the resulting feed solution with hexone,
nitrate ion de?ciency, contacting the resulting feed solu
scrubbing the resulting uranium containing organic phase
tion with hexone, scrubbing the resulting uranium and 30 with an aqueous solution of approximate composition
plutonium containing organic phase with an aqueous
0.5-3.0 molar aluminum nitrate, ‘0.05 molar ferrous sul
approximately 0.05-0.5 normal nitrate ionde?cient alu
famate, and 0.3-0.5 normal nitrate ion de?cient, separat
minum nitrate solution, separating the scrubbed uranium
ing the scrubbed uranium-containing organic phase from
and plutonium containing organic phase from the result
the resulting aqueous phase, and then stripping the sepa
ing ?ssion products-containing aqueous phase, contacting 3 C21 rated organic phase with an aqueous solution approxi
‘the separated organic phase with an aqueous, approxi
mately 0.05 normal in nitric acid.
mately 0.025-0.1 molar ferrous sulfamate solution, scrub
References Cited in the ?le of this patent
bing the resulting plutonium-containing aqueous phase
Iwith acidi?ed hexone, separating the resulting uranium
UNITED STATES PATENTS
containing organic phase from the resulting scrubbed 40 2,227,83 3
Hixson et al ____________ __ Jan. 7, 1941
plutonium-containing aqueous phase, contacting the sepa
OTHER REFERENCES
rated organic phase with an aqueous nitric acid solution
less than approximately one normal in nitric acid, thereby
Flagg et al.: Scienti?c American, vol. 187, No. 1, July
stripping said uranium from said organic phase.
4. The method of claim 3 wherein said feed solution is 45
adjusted to approximately 2.0 molar uranyl nitrate, 0.1
molar sodium dichromate and 0.2 normal nitrate ion de
?ciency; said aluminum nitrate scrub solution is approxi
1952, pp. 62-67, particularly pp. 65-66.
Proceedings of the International Conference on the
Peaceful Uses of Atomic Energy, held in Geneva Aug.
18-20, 1955, vol. 9, pp. 484-491; pub. by United Nations,
1956.
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