Note: Descriptions are shown in the official language in which they were submitted.
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PATENT
ROOM TEMPERATURE ELECTRODEPOSITION OF ACTINIDES FROM IONIC
SOLUTIONS
RELATED APPLICATION DATA
This application claims priority to U.S. Application Serial No. 13/764,282
filed February
11, 2013; and is related to U.S. Application Serial No. 13/268,138, filed 7
October 2011, which
are incorporated herein by reference.
GOVERNMENT FUNDING
This invention was made with government support under DE-AC07-051D14517 and DE-
FC07-061D14781 awarded by the Department of Energy. The government has certain
rights in
the invention.
BACKGROUND OF THE INVENTION
1. Field of the invention
The present invention relates to the field of deposition of metals, especially
electrodeposition of actinides, and especially the room temperature
electrodeposition of
lanthanides and actinides from ionic liquids.
2. Background of the Art
Reclaiming unspent nuclear materials, while separating and sequestering
fission products
is extremely important in the management of the growing stockpile of nuclear
waste. More
importantly, reclamation of the actinide metals is important for future safety
due to the possible
proliferation of weapons. (Morss, L. R.; Edelstein, N. M.; Fuger, J.; Katz, J.
J.; Kirby, H. W.;
Wolf, S. F.; Haire, R. G.; Burns, C. J.; Eisen, M. S. The Chemistry of the
Actinide and
Transactinide Elements; third.; Springer Netherlands, 2006) Finally,
reclamation of unused
uranium from nuclear fuel is of general importance for reuse in energy
processes and for the
production of target material to generate useful radio pharmaceutical species
for biological
applications. (Hofman, G. L.; Wiencek, T. C.; Wood, E. L.; Snelgrove, J. L.;
Suripto, A.;
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Nasution, H.; Amin, D. L.; Gogo, A. In 19th International Meeting on Reduced
Enrichment for
Research and Test Reactors; 1996.)
Typical electrochemical processes to recover uranium from spent nuclear fuel
result in
the accumulation of minor actinides (americium (Am) and curium (Cu)) and
transuranic
elements (plutonium (Pu) and neptunium (Np)). These accumulated elements
usually occur as
metal chlorides in the molten electrolytic salt. They must periodically be
removed from the
electrolyte for the fuel reprocessing to continue.
The simplest method to recover the target elements is via chemical or
electrochemical
reduction. Electrochemical reduction has two advantages over chemical
reduction. The first
advantage is that the site of reduction is localized to the cathode surface
forming a cathode
deposit affording easy removal from the process equipment. The second
advantage is that the use
of electrons as the reducing agent does not add to the waste volume.
Deposition of the
transuranic elements and minor actinides on a solid cathode is well-known.
Accompanying
anode reactions include the oxidation of chloride ions to chlorine gas,
oxidation of a sacrificial
alloy, and oxidation of metallic uranium or reduced light water reactor (LWR)
feed material.
U.S. Patent No. 7,267,754 (Willit) discloses an improved process and device
for the
recovery of the minor actinides and the transuranic elements (TRU's) from a
molten salt
electrolyte. The process involves placing the device, an electrically non-
conducting barrier
between an anode salt and a cathode salt. The porous barrier allows uranium to
diffuse between
the anode and cathode, yet slows the diffusion of uranium ions so as to cause
depletion of
uranium ions in the catholyte. This allows for the eventual preferential
deposition of transuranics
present in spent nuclear fuel such as Np, Pu, Am, Cm.
U.S. Patent No. 6,233,298 (Bowman) describes a subcritical reactor-like
apparatus for
treating nuclear wastes, the apparatus comprising a vessel having a shell and
an internal volume,
the internal volume housing graphite. The apparatus has means for introducing
a fluid medium
comprising molten salts and plutonium and minor actinide waste and/or fission
products. The
apparatus also has means for introducing neutrons into the internal volume
wherein absorption of
the neutrons after thermalization forms a processed fluid medium through
fission chain events
averaging approximately 10 fission events to approximately 100 fission events.
The apparatus
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has additional means for removing the processed fluid medium from the internal
volume. The
processed fluid medium typically has no usefulness for production of nuclear
weapons.
Uranium Separation Process, US Patent 3,030,176, Apr. 1962. This work outlines
the
dissolution of Uranium and the separation of species from fission products.
The work outlines
the use of molten salts in the separation. The advantage of our method is that
RTIL solutions are
ionic providing the same properties without the need for elevated temperatures
(500 ¨ 750 C)
that the molten salts require which reduces the production of unwanted gases
in the recovery
process.
Electroseparation of actinide and rare earth metals. US Patent 5,582,706, Dec
10, 1996.
The work outlines a pyrochemical process used to recover 99% of the
transmutable fission
materials. The process uses the electrochemical separation of the waste
component. Our method
does not require multiple paths or pyrochemical methods to achieve dissolution
or separation of
the fission products.
Actinide Dissolution. US patent 5,205,999, Apr. 27, 1993. The work documents
the
dissolution of the actinide and lanthanide species in aqueous solution between
pH 5.5 to 10
utilizing complexing agents. Our methods are conducted under similar
conditions in room
temperature ionic liquid. The same solution is used for the electrochemical
separation and
deposition of actinide species. Our method does not require complexing agents,
it is performed
in non-aqueous solution, and the direct dissolution is achieved in the same
solvent system used
for electrodepositon.
Magnesium Reduction of Uranium Fluoride in Molten Salts. US patent 4,552,588,
Nov.
12, 1985. The work documents the use of Mg molten salts in the reduction of
UF4 to U metal.
The temperatures required for the process are on the order of 1000 degrees.
There are inherent
dangers associated with molten salts at high temperatures that are eliminated
when RTIL
solutions are used.
To date the PUREX process is the most widely utilized methods for the
reclamation of
actinides (Uranium and Plutonium) from partially spent nuclear materials.
PUREX is an
acronym standing for Plutonium - URanium EXtraction ¨the standard aqueous
nuclear
reprocessing method for the recovery of uranium and plutonium from used
nuclear fuel. It is
based on liquid-liquid extraction ion-exchange. The PUREX process was invented
by Herbert H.
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Anderson and Lamed B. Aspreyas part of the Manhattan Project. Their U.S.
Patent No.
2,924,506,"Solvent Extraction Process for Plutonium" filed in 1947, mentions
tributyl phosphate
as the major reactant which accomplishes the bulk of the chemical extraction.
The method utilizes a complexing agent, tri-n-butylphosphate (TBP) and organic
solvent
such as kerosene or n-dodecane in the extraction and reclamation process.
Modifications to the
process have been primarily focused on developing new complexing agents or
using different
solvents for extraction. More recently the RTIL solutions have been examined
as an alternative
to more volatile organic diluents using tricaprylmethylammonium thiosalicylate
as the
complexing agent in the extraction of U into RTIL solution. Srncik, M.;
Kogelnig, D.;
Stojanovic, A.; Koerner, W.; Krachler, R.; Wallner, G. Uranium extraction from
aqueous
solutions by ionic liquids, Applied Radiation and Isotopes (2009), 67(12),
2146-2149. The
added benefit of RTIL solutions is that it can be used in the direct
electrochemical deposition of
lanthanide or actinides species due the large potential window afforded by the
non-aqueous
system.
At present the accepted electrochemical method utilized to obtain uranium
metal is based
on molten salt eutectic system. (Iizuka, M.; Koyama, T.; Kondo, N.; Fujita,
R.; Tanaka, H.
Journal of Nuclear Materials 1997, 247, 183-190. Kim, K. R.; Bae, J. D.; Park,
B. G.; Ahn, D.
H.; Paek, S.; Kwon, S. W.; Shim, J. B.; Kim, S. H.; Lee, H. S.; Kim, E. H.;
Hwang, I. S. J
Radioanal Nucl Chem 2009, 280, 401-404. Koyama, T.; Iizuka, M.; Shoji, Y.;
Fujita, R.; Tanaka,
H.; Kobayashi, T.; Tokiwai, M. Journal of Nuclear Science and Technology 1997,
34, 384-393.
Internationally there are two well developed molten salts processes for the
reprocessing/waste conditioning of irradiated nuclear fuel. A process
developed by the
Dimitrovgrad SSC¨RIAR process uses high temperature (1000K) eutectic molten
salt mixtures
as solvents for the fuel and also as electrolyte systems. In this Russian
system the solvent is
typically an eutectic mixture of NaCl/KC1 or CsC1/KC1. The process uses
chemical oxidants
(chlorine and oxygen gases) to react with powdered UO2 fuel, or mixtures of
UO2 and Pu02, to
form higher oxidation state compounds such as UO2C12 which are soluble in the
molten salt. At
the cathode the uranium and, if applicable, plutonium compounds are reduced to
UO2 or UO2¨
Pu02, which form crystalline deposits. However, after a period of use the
molten salt becomes
loaded with fission products which not only begin to affect the quality of the
product, but also
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result in too much heat generation within the salt. These fission products are
commonly, but not
exclusively, highly active lanthanide or actinide elements which may need to
be isolated in a
suitable form for immobilisation as a waste.
In the process developed by Argonne National Laboratory (ANL) in the USA,
molten
LiC1/KC1 eutectic mixtures containing some UC13 are generally used, rather
than systems
containing sodium or caesium salts, and a high temperature (around 773K) is
again employed.
However, single salts, such as LiC1, are suitable if higher temperatures are
required, for example
in the electrochemical reduction of fuel oxides. The process treats the spent
nuclear fuel by
flowing a current to oxidize a uranium anode and form uranium ions in the
molten salt
electrolyte. At the cathode the uranium is reduced and deposited as uranium
metal. The ANL
process is, unfortunately, a batch process, since the uranium is collected in
a receptacle at the
bottom of the apparatus, requiring that the process is interrupted in order
that the receptacle may
be withdrawn and the product recovered. In addition, the operation of the
process is
mechanically intense, involving the use of rotating anodes which are designed
to scrape the
product off the cathodes; difficulties are encountered on occasions due to the
seizure of this
mechanism.
While these methods have been utilized to produce U metal, it is not without
flaws. From
an engineering standpoint, the high temperatures needed for a molten salt
system create safety
and cost issues for the vessel material fabrication. (Avallone, E.;
Baumeister, T.; Sadegh, A.
Marks' Standard handbook for mechanical Engineers; 1 lth ed.; Mc-Graw Hill
Professional,
2006.Creep & Fracture in High Temperature Components: Design & Life Assessment
Issues;
Shibli, I.; Holdsworth, S.; Merckling, G., Eds.; DESTech Publications, Inc.,
2005.). In addition,
gas evolution is problematic due to environmental concerns and the safety of
the workers. The
second method is based on the synthesis ofUF4 using HF gas.( Pushparaja;
Poplit, K.; Kher, R.;
Iyer, M. Radiation protection dosimetry 1992, 42, 301-305.) The process is
expensive and
dangerous process due to the health hazards and corrosive nature of
hydrofluoric acid. In
addition, reduction of the UF4 to metal using plasma and hydrogen is
complicated by
disproportionation and production of UF3 limiting the overall metal
conversion.
All references cited herein are incorporated in their entirety by reference.
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SUMMARY OF THE INVENTION
The present invention relates to a method for the electrochemical deposition
of an
actinide or lanthanide with at least steps of: providing an actinide ion in a
room temperature ionic
liquid to form an actinide rich liquid composition; providing an electrode and
a cathode within
the actinide rich liquid composition; and at temperatures below 30 C, applying
a potential such
that current passes between the electrode and cathode to deposit actinide
metal on the cathode.
BRIEF DESCRIPTION OF THE FIGURES
Figure 1 shows a graphic representation of the Cyclic Voltammetric response of
an Au
electrodein RTIL (dashed line) and RTIL solution containing U(TFSI)3 (solid
line).
Figure 2 shows a photomicrograph (scanning electron micrographs) of Top: SEM
image of Au
surface prior to deposition. Bottom: SEM image of U deposited on Au from RTIL
solution containing U(TFSI)3.
Figure 3 shows a graph of an Energy dispersive spectra for U deposits on an Au
electrode from
RTIL solution containing U(TFSI)3.
Figure 4 shows a graphic representation of Powder XRD fit for uranium deposits
from U(TFSI)3
on a gold electrode.
DETAILED DESCRIPTION OF THE INVENTION
The present disclosure provides encompasses methods of introducing varying f-
species
into a Room Temperature Ionic Liquid (RTIL) using extraction or direct
dissolution.
Introduction of an actinide or lanthanide without organic diluent or a
secondary complexing
agent into the RTIL solvent is novel and has not yet been demonstrated. The
direct dissolution of
uranium complexes and the potential dependent deposition of these species in
metal form has not
been present in previous literature. While direct addition of certain
lanthanide and actinide
species has been documented; very little information is available regarding
the potential
mediated lectrodeposition of the corresponding f-metals from the RTIL solvent
at room
temperature. For example, the deposition and subsequent identification of
metallic uranium at
room temperatures has not been published in the literature to date.
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Room temperature ionic liquids (RTILs) are a potential solution to molten
salts because
they have similar electrochemical properties without the need for elevated
temperature. The
large potential window of RTIL solutions is useful for electrochemical
reduction of oxidized
actinides and lanthanides, they have negligible vapor pressures, and are
stable chemically even at
elevated temperatures. (Reddy, R. G. JPED 2006, 27, 210-211. Cocalia, V. A.;
Gutowski, K. E.;
Rogers, R. D. Coordination Chemistry Reviews 2006, 250, 755-764. Earle, M. J.;
Seddon, K. R.
Pure and Applied Chemistry 2000, 72, 1391-1398.). Finally the thermodynamic
driving force
for the reduction of the species can be controlled precisely minimizing side
reactions and
disproportionation common to plasma based reduction of actinide halide
complexes. In this
specification, the term "ionic liquid" essentially refers to a salt which
melts at a relatively low
temperature. For example, the electrochemical reactions in RTIL can be
conducted at room
temperature or moderately elevated temperatures in the range of 30 ¨ 200 C
without significant
degradation of the ionic solvent. Ionic liquids free of molecular solvents
were first disclosed by
Hurley and Wier in a series of U.S. Pat. Nos. (2,446,331, 2,446,349,
2,446,350). Common
features of ionic liquids include a near zero vapor pressure at room
temperature, a high solvation
capacity and a large liquid range (for instance, of the order of 300 C.).
Known ionic liquids
include aluminium(III) chloride in combination with an imidazolium halide, a
pyridinium halide
or a phosphonium halide. Examples include 1-ethyl-3-methylimidazolium
chloride, N-
butylpyridinium chloride and tetrabutylphosphonium chloride. An example of a
known ionic
liquid system is a mixture of 1-ethy1-3-methylimidazolium chloride and
aluminium (III) chloride.
The RTIL system of the present technology may include an asymmetric organic
cation
and a large anion that can both be varied to influence the solution properties
including solubility,
viscosity, and the overall potential window for electrochemical experiments.
(Earle, M. J.;
Seddon, K. R. Pure and Applied Chemistry 2000, 72, 1391-1398. Buzzeo, M. C.;
Evans, R. G.;
Compton, R. G. ChemPhysChem 2004, 5). In this work, the anion selected was n-
Bis(trifluoromethansulfonylimide) (TFSI), and the cation was trimethyl-n-butyl
amine. The
combination of this pair allows for a low melting point liquid with high ionic
conductivity. In
addition, the potential window for this solvent system is on the order of six
volts encompassing
negative potentials for the reduction of both lanthanides and actinides to
metal. Solubility can be
an issue when trying to introduce species into the RTIL. While solubility can
be influenced
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using different combinations of cation/anion pairs, the combinatorial approach
required to
identify the RTIL species is not feasible due to the sheer magnitude of pairs
that exist and the
inherent cost. Therefore, forming complexes with anions common to the RTIL
were specifically
targeted to enhance solubility of the species in RTIL.
Previous work suggests that the electrochemical deposition of Uranium metal is
possible
under appropriate molten solvent conditions. For example, Uranium metal
deposits were
successfully obtained from U(III) and U(IV) complexes in molten salt
systems.16 For comparison
the uranium complexes U(TFSI)3 and U(TFSI)4 were prepared in our laboratory
for the
electrochemical studies using RTIL. However, we will focus on the U(TFSI)3
system. All
experiments were performed in an Argon evacuated glove box to minimize the
formation of
oxides after reduction of the uranium TFSI complexes in RTIL. The complexes
directly dissolve
in the RTIL after addition.
Uranium metal can be electrochemical deposited from room temperature ionic
liquid
(RTIL), tri-methyl-n-butyl ammonium n-bis(trifluoromethansulfonylimide),
[Me3N13u][TFSI]
providing an alternative non-aqueous system for the extraction and reclamation
of actinides from
reprocessed fuel materials. Furthermore, deposition of U metal is achieved
using TFSI
complexes of U(III) and U(IV) containing the anion common to the RTIL. The
goal was to
produce TFSI complexes of uranium to ensure solubility of the species in the
ionic liquid. The
methods outlined provide a first measure of U metal deposition using Uranium
complexes with
different oxidation states from RTIL solution at room temperature.
The US Argonne National Laboratory developed a new apparatus called Plannar
electrode Electrorefiner (PEER) at http://www.cmt.anl.gov. The apparatus is
designed to deposit
an anode including a metallic fuel in the middle and a plurality of cathodes
therearound and
operate an electrolytic reaction. After a certain time passes, the
electrodeposites are attached on
the cathode and a porous ceramic plate is moved in a vertical direction to
scrap out the cathode
electrodeposites. In general, when an electrorefining process is carried out,
the density of a
current applied to an electrode relates to an electrodeposition rate in a
cathode and a sticking
coefficient. As the current density is increased, a lot of uranium can be
electrodeposited for a
short time when it comes to the electrolytic rate. The sticking coefficient is
defined as the
amount of the electrodeposites stuck to a cathode surface to the amount of
uranium metal
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transmitted to the cathode. Therefore, if the current density is increased
using the electrode, the
electrolytic rate is increased to decrease the sticking coefficient. The
magnitude of the current
density applied to the apparatus for an electrorefining or electrodeposition
according to the
present invention depends on the content of an allowable electrodeposite,
preferably the current
density of which the sticking coefficient is 0%. The current density of which
the sticking
coefficient is 0% may be defined experimentally. For example, a current
density greater than of
between the current used during step c) to cause electrodeposition is in the
range of between
10[Lamps and 500 [tamps /cm2 and preferably between 50[Lamps and 150 or 200
[tamps /cm2 is a
range that can be conveniently applied in one embodiment of the present
invention using a single
carbon rod as a cathode.(While the above paragraph represents the goal of this
research, we have
not quantified the terms that you have defined. I do not disagree with the
numbers quoted.
However, we have not optimized the process and this must be done to define
these parameters
for the system we are using. It may be very different than the system used to
define these terms
from Los Alamos. In addition we used applied potential and the current that
resulted was ¨10
microA/cm2)
The electrochemical response for U(TFSI)3 (solid line) is presented in Figure
1 with the
corresponding background (dashed line) for the RTIL. The cyclic voltammetric
response for
U(TFSI)3 is for the 10th cycle. There is a much higher current density due to
the increased
surface area associated with increased surface deposits of Uranium. Sequential
cycle results in
an increase in current density as the surface deposit increases increasing the
overall surface area
on the electrode (not shown). A voltammetric reduction wave is observed in the
negative
potential scan at --1.25 V consistent with the deposition of U(0) on the
electrode surface. The
reverse scan shows a voltammetric wave at ¨0.75 which can be attributed to the
combined
oxidation of U(III) to U(IV) and the partial oxidation of the U deposits.
However, there is a net
increase in surface deposition of uranium after each voltammetric cycle. The
electrochemical
deposition was achieved using multiple techniques include cyclic voltammetry
and constant
potential methods. For the constant potential methods deposition was conducted
at/or more
negative than -2.0 V. Dark grey deposits were obtained on the electrode
surface indicative of U
metal deposition.
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Scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS)
analysis
was used to evaluate the deposit and provide information regarding the
speciation. The electrode
was protected from air during transportation by sealing the sample argon
evacuated container
immediately prior to placement in the SEM. The SEM image of a clean gold
electrode (top) and
the deposited electrode (bottom) are shown in Figure 2. The Au surface is
clearly visible in the
SEM image for the deposited electrode. The surface deposits were examined at
eleven sights
using EDS, Figure 3. The EDS spectrum has bands characteristic of the U
deposits, with some
residual S from the RTIL. The deposits are sufficiently thick that the
contribution of Au to the
EDS spectrum is not observed. In addition, the uranium deposits were observed
with no
detectable oxygen in the EDS response. The results confirm that the
electrochemical deposition
of U metal from U(TFSI)3 complex is feasible from RTIL solutions.
Secondary analysis of the deposits was conducted following SEM-EDS analysis
using
powder x-ray diffraction (XRD) for deposits from U(TFSI)3 in RTIL, Figure 4.
This sample was
also protected from air in a similar manner as for the SEM analysis, including
sealing the XRD
sample container. The crystallographic phase of the uranium deposit was
evaluated using
powder XRD. Uranium can be found in three phases: alpha, beta, and gamma at
temperatures
below 1135 C. The alpha phase is the predominant form at room temperature.
Although the
penetration of the source is such that the Au electrode is the predominant
species in the powder
XRD, alpha uranium metal is observed. The XRD response confirms the EDS
analysis that
uranium metal is deposited on the Au electrode.
The studies outlined demonstrate that uranium metal deposition was achieved at
room
temperature from ionic liquid containing trimethyl-n-butyl amine cation with
the TFSI anion.
Furthermore, the nature of the deposit was analyzed using both SEM-EDS and
powder XRD.
The deposition of alpha uranium metal was confirmed on the Au electrode
surface. The results
suggest that actinide deposition from RTIL solutions my useful in replacing
alternative methods
for obtaining uranium metal including chemical vapor deposition with plasma
reduction and
deposition from molten salt systems.
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PROPHETIC EXAMPLE
The deposition of U has been achieved using three different species: U(TFSI)3,
U(TFSI)4,
and UI3(THF)2. Each sample was prepared using 4 ml of RTIL solution with ¨10mg
of total U
content. The complexes were dissolved directly into the ionic liquid with
simple mixing.
Similar methods were used for the dissolution and deposition of Sm metal using
Sm(TFSI)3 into
the RTIL. The potential dependent deposition of U from RTIL solutions
containing the
complexes was conducted using a three electrode cell containing a cathode
(Glassy Carbon disk
or Au sheet) with a Pt counter electrode and a Ag/Ag ' reference electrode
filled with 0.1 M
AgNO3 in RTIL. Current versus time plots were obtained using an applied
potential for each
experiment of -2.2 V vs. NHE. The deposition was conducted over a 24 hour
period without
mixing. Under these conditions 6-8 mg/cm2 of U metal was deposited at the gold
cathode using
normal diffusion with a measured current density on the order of 10 ilA/cm2.
The gold cathode
was in the form of a sheet (1 cm2) that was transferred directly to do both
the TEM and XRD
analysis of the deposits. Similar deposition was achieved using glassy carbon
disk electrodes as
the cathode. Cyclic voltammetric techniques were also utilized to deposit U
metal at the cathode
as shown in Figure 3.
The above prophetic example is based on actual experiments, however, the
example has
been broadened without determining the efficiency of deposition of U. However,
these
experiments can determine the efficiency of the processes using different U
complex materials.)
The applications of this process are far reaching. It could simply be the
reclamation of unused U
from used fuel. It could also be the production of U targets for use in cross-
section
measurements. Finally, it could be used in the dissolution and reclamation of
Tc99, a
radiopharmaceutical that is produced during U fission processes.
Other variants, alternatives and substitutions can be provided by one skilled
in the art
within the framework of the generic invention described herein.
The sequential dissolution of 0.1 gram quantities of U308 (s) in IL containing
0.1 M
HTFSI using ozone was conducted. The dissolution was examined at ambient
temperature using
an ozone stream of 1-2 wt % at rate of 450 cc/min in an air stream of 18-20%
oxygen. Complete
dissolution of the U308 solid occurred within twenty-four hours after each 0.1
g addition without
adding any additional water or acid to the system. Increases in the absorbance
for the soluble
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uranyl ion are observed for increasing concentrations of soluble U308.
Previously the dissolution
of U308 was not achieved in task specific ionic liquid, [CH3WCOOFI][TFSI] at
temperatures
below 473 K and only limited solubility (¨ 0.5 ¨ 0.75 wt %) was achieved at
elevated
temperatures. For comparison, the solubility achieved for the dissolution of
U308 using ozone is
3% by weight in [Me3N13u][TFSI] containing 0.1 M HTFSI. The measured value
does not
represent the maximum solubility of U308 because each addition dissolved
completely and
saturation was not achieved in the IL. The electrochemical deposition of UO2
(s) was also
achieved from the 3% solution of soluble U308 on an electrode surface
providing a mass density
of 10 mg/cm2. The example provided highlights the dissolution and recovery of
uranium oxide
from ionic liquid.
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