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\DOI{10.5802/crchim.320}
\datereceived{2023-12-04}
\daterevised{2024-05-07}
\dateaccepted{2024-05-22}
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\begin{document}

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%\makeatletter
%\def\TITREspecial{\relax}
%\def\cdr@specialtitle@english{GDR Prom\'eth\'ee - French Research Network on \textit{Hydrometallurgical Processes for Primary and Secondary Resources}}
%\def\cdr@specialtitle@french{GDR Prom\'eth\'ee - Groupement de Recherche sur les Proc\'ed\'es Hydrom\'etallurgiques pour la Gestion Int\'egr\'ee des Ressources Primaires ...}
%\makeatother

\title{Back-extraction of major actinides from organic phase through
precipitation}

\alttitle{D\'{e}sextraction pr\'{e}cipitante d'actinides \`{a} partir
d'une phase organique}

\author{\firstname{Julie} \lastname{R\'{e}aux-Durain}}
\address{ICSM, CEA, CNRS, ENCSM, University of Montpellier, Marcoule, France}

\author{\firstname{Boushra} \lastname{Mortada}}
\addressSameAs{1}{ICSM, CEA, CNRS, ENCSM, University of Montpellier, Marcoule, France}

\author{\firstname{Murielle} \lastname{Bertrand}\CDRorcid{0000-0002-8013-5911}}
\address{CEA, DES, ISEC, DMRC, BP 17171, 30207 Bagnols-sur-C\`{e}ze Cedex, France}

\author{\firstname{J\'{e}r\^{o}me} \lastname{Maynadi\'{e}}}
\addressSameAs{1}{ICSM, CEA, CNRS, ENCSM, University of Montpellier, Marcoule, France}

\author{\firstname{Micha\"{e}l} \lastname{Carboni}\CDRorcid{0000-0003-0828-9067}}
\addressSameAs{1}{ICSM, CEA, CNRS, ENCSM, University of Montpellier, Marcoule, France}

\author{\firstname{Daniel} \lastname{Meyer}}
\addressSameAs{1}{ICSM, CEA, CNRS, ENCSM, University of Montpellier, Marcoule, France}

\author{\firstname{Damien} \lastname{Bourgeois}\CDRorcid{0000-0002-0762-3476}\IsCorresp}
\addressSameAs{1}{ICSM, CEA, CNRS, ENCSM, University of Montpellier, Marcoule, France}
\email[D. Bourgeois]{damien.bourgeois@umontpellier.fr}

\shortrunauthors

\keywords{\kwd{Solvent extraction}
\kwd{Stripping}
\kwd{Precipitation}
\kwd{Uranium}
\kwd{Thorium}
\kwd{Oxalic acid}
\kwd{Tributylphosphate}}

\altkeywords{\kwd{Uranium}
\kwd{Thorium}
\kwd{Extraction}
\kwd{Phosphate de tributyle}
\kwd{Acide oxalique}}

\begin{abstract}
Recycling 
waste nuclear fuel through the production of mixed oxide
(MOx) fuel allows for sustainable management of natural nuclear
resources. With the aim of reducing the number of steps required for
recycling U and Pu, the possibility of precipitating them directly from
the organic phase (also referred to as precipitation stripping) using
oxalic acid, is investigated herein. For practical reasons,
proof-of-concept studies were performed using U(VI) and Th(IV), with
the latter serving as a first model for Pu(IV). The results
obtained show that, by controlling the concentrations of oxalic and
nitric acids in the precipitation medium, total precipitation of Th and
partial precipitation of U from the organic phase are attained. The
actinides are precipitated in the form of oxalates, with the complete
absence of impurities. Furthermore, the precipitates obtained can then
be totally separated through washing with water, as proven by X-ray
diffraction (XRD) and energy-dispersive X-ray (EDX) analyses. This
allows control over the U:Th mass ratio in the final product. This work
demonstrates the effectiveness of precipitation stripping using oxalic
acid for the stripping of Th and U present in the organic phase,
eliminating the need for reducing agents classically employed in the
PUREX process.
\end{abstract}

\begin{altabstract}
Le recyclage des d\'{e}chets de combustible nucl\'{e}aire par la
production d'oxyde mixte (MOx) permet une gestion durable des
ressources naturelles nucl\'{e}aires. Dans le but de r\'{e}duire le
nombre d'\'{e}tapes n\'{e}cessaires au recyclage de l'U et du Pu, la
possibilit\'{e} de les pr\'{e}cipiter directement \`{a} partir d'une
phase organique (op\'{e}ration \'{e}galement appel\'{e}e
d\'{e}sextraction pr\'{e}cipitante) en utilisant de l'acide oxalique a
\'{e}t\'{e} \'{e}tudi\'{e}e. Pour des raisons pratiques, des \'{e}tudes
de validation de concept ont \'{e}t\'{e} effectu\'{e}es en utilisant
U(VI) et Th(IV), ce dernier \'{e}tant utilis\'{e} comme mod\`{e}le
simplifi\'{e} du Pu(IV). Les r\'{e}sultats obtenus r\'{e}v\`{e}lent
qu'en contr\^{o}lant les concentrations en acides oxalique et nitrique
dans le milieu, la pr\'{e}cipitation totale de U et de Th \`{a} partir
de la phase organique est r\'{e}alis\'{e}e. Les actinides sont
pr\'{e}cipit\'{e}s sous forme d'oxalates, sans aucune autre
impuret\'{e}, comme le prouvent les analyses de diffraction des rayons
X (XRD) et de dispersion d'\'{e}nergie (EDX). En outre, les
pr\'{e}cipit\'{e}s obtenus peuvent ensuite \^{e}tre totalement
s\'{e}par\'{e}s par lavage \`{a} l'eau, ce qui permet de contr\^{o}ler
le ratio \mbox{U : Th} dans le solide final. Ce travail montre clairement
l'efficacit\'{e} de la d\'{e}sextraction par pr\'{e}cipitation \`{a}
l'aide d'acide oxalique pour la r\'{e}cup\'{e}ration de l'U et du Th
pr\'{e}sents dans une phase organique, et ouvre la voie vers la
formation de pr\'{e}curseurs de mat\'{e}riaux mod\`{e}les MOx d'un
ratio \mbox{U : Th} compl\`{e}tement contr\^{o}l\'{e}.
\end{altabstract}

\maketitle

\vspace*{1pt}

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\let\rmmu\upmu

\section{Introduction}\label{sec1}
The operation of the reactors of the French nuclear power plants is
based on a fuel composed of natural uranium oxide, enriched in isotope
235, and for some reactors, of a mixture of uranium and plutonium
oxides, known as MOx (Mixed Oxide) fuel. The production of MOx fuel
enables recycling both uranium and plutonium present in waste nuclear
fuel, and thus a sustainable management of natural nuclear 
resources~\cite{1}. 

The treatment of spent nuclear fuel is currently carried out in France
following the PUREX (Plutonium, Uranium, Reduction, Extraction) 
process~\cite{2}. This consists in dissolving the spent nuclear fuel in
nitric acid HNO$_{3}$, followed by the separation of uranium(VI) and
plutonium(IV) from the aqueous medium through liquid--liquid extraction
using tributylphosphate (TBP) as an extracting  molecule~\cite{2,3}.
Afterwards, two back-extraction (stripping) steps toward aqueous phases
are performed to separate U from Pu, with the stripping of Pu relying
on the reduction of Pu(IV) to Pu(III). Pu(III) is then precipitated
from the aqueous phase using oxalic acid (C$_{2}${O}$_{4}$H$_{2}$).
U(VI) is precipitated into uranium(VI) oxalate using the same
technique. The solids obtained are then calcined so as to form the
corresponding oxides, mixed (in controlled proportions) to form MOx and
shaped into pellets to be used as fuel for the nuclear reactors.
Inspired by the PUREX process, the  COEX{\texttrademark}  process was
developed by the CEA (\textit{Commissariat \`{a} l'Energie Atomique et
aux Energies Alternatives}) with the aim of decreasing the number of
steps during uranium and plutonium recycling, in addition to preventing
the generation of pure-plutonium effluents, which may be used in
non-civil  applications~\cite{4,5}. Similar to the PUREX process, the 
COEX{\texttrademark} process involves in the first two steps the
dissolution of waste nuclear fuel in nitric acid followed by extraction
of both uranium and plutonium through liquid--liquid extraction using
also TBP. The back-extraction step towards the aqueous phase has been
optimized in the  COEX{\texttrademark} process, so as to generate, in
addition to an aqueous flux containing only uranium, another one
containing both uranium and plutonium  (U:Pu ${<}$ 20\%)~\cite{1}. Once
uranium and plutonium from the mixed aqueous flux have been purified
from minor impurities by liquid--liquid extraction, this flux can be
directly used for the production of MOx fuel. This is accomplished
through the oxalic coprecipitation of uranium(IV) and plutonium(III)
from the aqueous  phase~\cite{6}, followed by a calcination step, which
results in the formation of a mixed oxide (U,Pu)O$_{2}$ powder.
Whether in the PUREX or  COEX{\texttrademark} processes, oxalic acid
(co)precipitation is carried out in aqueous medium, after a
back-extraction step of Pu involving Pu(III). Thus, even though these
two recycling methods are efficient in the production of MOx, they
still involve numerous steps, especially due to the need to proceed to
Pu(IV) reduction into Pu(III). The latter reduction is usually
performed using U(IV) generated using hydrazine as a reducing agent, a
carcinogenic substance. 

Back-extraction of metallic elements from an organic phase through
precipitation, also known as precipitation stripping, has been reported
in the  literature~\cite{7,8,9,10,11}. This process combines both the
concentration of metal ions in the aqueous phase (stripping) and
precipitation stages when bringing the metal-containing organic
solution into contact with the aqueous solution containing the
precipitating agent. Most of these examples rely on oxalic acid use,
and have been applied to the separation and purification of heavy
metals and lanthanides. The process results in the formation of metal
oxalate precipitates in the aqueous (C$_{2}${O}$_{4}${H}$_{2}$) phase,
from which they can be collected by filtration. The extraction of
plutonium (IV) from a plutonium-containing calcium fluoride slag issued
from the production of plutonium metal was realized through direct
oxalic precipitation from the organic  phase~\cite{12}. Plutonium was
first extracted in an organic solution containing TBP at 20\% (v/v) in
kerosene. The plutonium-charged organic phase was then put in contact
with an aqueous solution of 0.25~mol/L C$_{2}${O}$_{4}${H}$_{2}$ in
HNO$_{3}$ (1.5 to 3.0~mol/L). After  30~min of stirring, plutonium was
precipitated in the aqueous phase in the form of a wet cake of
plutonium  oxalate~\cite{12}. Later on a process for
recovering/separating plutonium(IV) from a mixture of plutonium(IV) and
uranium(VI) or americium(III) was reported in the literature, through
oxalic precipitation from an organic phase of TBP at\unskip\break 15--30\% (v/v)  in
$n$-dodecane~\cite{13}. This resulted in the precipitation of plutonium
in the form of plutonium oxalate in the aqueous phase, whereas
uranium(VI) and americium(III) remained mainly concentrated in the
organic solution. 

To our knowledge, the oxalic coprecipitation of uranium and plutonium
from the organic phase has not yet been reported in the literature.
Therefore, with the aim of reprocessing uranium and plutonium issued
from spent nuclear fuel through the production of MOx, we have studied
the possibility of coprecipitating both actinides directly from the
organic phase using oxalic acid as a precipitating agent through
precipitation stripping. This approach aims to reduce the number of
steps required for recycling U and Pu compared to the PUREX and 
COEX{\texttrademark} processes, by avoiding the back-extraction into
the aqueous phase through reduction of Pu(IV) into Pu(III). The key
points presented in this study are (1)~the efficiency of the
precipitation of both actinide(IV) and actinide(VI), and (2)~the
possibility of controlling the ratio of both metals in the precipitate
obtained. The latter objective was motivated by the need to develop
non-proliferating spent fuel reprocessing sequences, which avoid
isolation of pure plutonium, as detailed in one of our previous 
studies~\cite{14}. This proof-of-concept study was carried out using
thorium(IV) as a substitute for plutonium(IV). The solids obtained were
characterized to assess the mechanism of precipitation and their
suitability for use in the further reprocessing sequence. The presence
of an unwanted by-product was evidenced in the case of uranium(VI)
precipitation stripping, and the experimental conditions were optimized
to inhibit its occurrence. 

\vspace*{-2pt}

\section{Experimental section}\label{sec2}

\vspace*{-2pt}

All the reagents were used as received and without any further
modification. Uranyl nitrate hexahydrate
$\mathrm{UO}_{2}(\mathrm{NO}_{3})_{2}{\cdot}6\mathrm{H}_{2}\mathrm{O}$
and $\mathrm{Th}(\mathrm{NO}_{3})_{4}{\cdot}5\mathrm{H}_{2}\mathrm{O}$
were supplied by the CEA.  Tri-$n$-butylphosphate (TBP, 97\%
pure), $n$-dodecane (${<}$99\% pure) and oxalic acid dihydrate were
purchased from Sigma aldrich/Merck. Powder XRD (PXRD) patterns were
obtained on a Bruker D8 Advance diffractometer, using CuK${\upalpha}$
radiation (${\lambda}= 1.5418$~\AA{}). Scanning electron microscopy
(SEM) images were acquired using a FEI Quanta 200 environmental
scanning electron microscope, the morphology of the samples was
recorded with a secondary electron (SE) detector, whereas a
backscattered electron detector (BSED) was used to detect the presence
of the different elements constituting the precipitates through energy
dispersive X-ray (EDX) analyses. Fourier transform infrared (FT-IR)
spectroscopy analyses were performed on a IR-FT100 spectrometer from
Perkin Elmer, equipped with an MIR source, a DTGS detector and an
attenuated total reflectance (ATR) measurement mode. Analyses were
carried out by placing a drop of the liquid or a pinch of the solid on
the surface of the diamond analyzer. Thermogravimetric and
thermodifferential analyses coupled with mass spectrometry
(TGA-TDA-MS) were carried out on a SETSYS EVO 16 machine supplied by
Setaram and coupled with a Hiden QGA300 mass spectrometer for gas
analysis. 

\vspace*{-2pt}

\subsection{Liquid--liquid extraction: obtaining an\newline organic phase
highly charged in actinides}

\vspace*{-2pt}

The general liquid--liquid extraction protocol involved an aqueous
solution of 
$\mathrm{UO}_{2}(\mathrm{NO}_{3})_{2}{\cdot}6\mathrm{H}_{2}\mathrm{O}$
(0.326~M) or a mixture of
$\mathrm{UO}_{2}(\mathrm{NO}_{3})_{2}{\cdot}6\mathrm{H}_{2}\mathrm{O}$
(0.326~M) and
$\mathrm{Th}(\mathrm{NO}_{3})_{4}{\cdot}5\mathrm{H}_{2}\mathrm{O}$
(0.18~M) in HNO$_{3}$ 6~M. An organic solution of TBP (1~M) in 
$n$-dodecane was prepared and pre-equilibrated
with 6~M HNO$_{3}$ by
mixing the two solutions in an ampoule, shaken manually for  10~min.
Afterward, both phases were separated\unskip\break
(A:O volume ratio equal to 5:1),
and the aqueous phase was discarded. The organic solution of TBP and
the aqueous solution containing U(VI) and Th(IV) were introduced into
an Eppendorf tube and shaken in an Eppendorf Thermomixer at 
20~\textdegree C and 100~rpm during 10~min. At the end of the extraction
step, the organic solution, loaded in actinides, was separated from the
aqueous phase. The former was stored for use in organic-phase oxalic
precipitation. 

\vspace*{-1pt}

\subsection{Precipitation stripping of uranium and\newline thorium}

\vspace*{-1pt}

An aqueous C$_{2}$O$_{4}$H$_{2}$ solution in HNO$_{3}$ (see details in
the Results and discussion section for the concentrations of
C$_{2}$O$_{4}$H$_{2}$ and HNO$_{3}$, volume 500~$\rmmu$L) was
introduced into a 5~mL glass vial equipped with a magnetic stirrer.
This was followed by the addition of the organic solution recovered
from liquid--liquid extraction in a dropwise manner 
(500~$\rmmu$L) while stirring at 500 rpm at
$20\pm2~\text{\textdegree}\mathrm{C}$. Stirring was maintained for 
1--2~h (see the Results and discussion section for the precise stirring time
for the different experiments), after which the mixture was
centrifuged, and the aqueous and organic phases, in addition to the
solid precipitate, were collected. The liquid phases were stored in
Eppendorf tubes for further ICP analyses (see below). The solid
precipitate was washed with  $n$-dodecane, then with 
$n$-heptane, and then dried at  40~\textdegree C for a few hours. The
precipitation yield was determined from the mass balance after analysis
of both remaining liquid aqueous and organic phases.

\vspace*{-1pt}

\subsection{Quantification of uranium and thorium in solutions}

\vspace*{-1pt}

Quantification of metals in solution was performed using inductively
coupled plasma atomic-emission spectrometry  (ICP-AES) with a Spectro
Acros apparatus from Ametek. Aqueous phases were directly analyzed,
using an aliquot directly diluted in a matrix solution consisting of
1\% aqueous HNO$_{3}$ and 1\% aqueous HCl, with a minimum HNO$_{3}$:HCl
volume ratio of 9:1. In order to determine the concentration of the
actinides remaining in the organic phase after precipitation, these
were back-extracted by a solution of HNO$_{3}$ 0.01~M (A:O volume ratio
equal to 100:1). The organic and aqueous solutions were \mbox{introduced} into
an Eppendorf tube and shaken at  20~\textdegree C and 1000~rpm for
10~min. The aqueous phase was then recovered and diluted in a matrix
solution consisting of 1\% aqueous HNO$_{3}$ and 1\% aqueous HCl with a
HNO$_{3}$:HCl volume ratio of 9:1.

\section{Results and discussion}\label{sec3}
The experimental studies were conducted on an organic phase designed in
order to mimic the extraction phase obtained during the first stage of
the PUREX process~\cite{1}. In the PUREX process, after spent fuel
dissolution in nitric acid, uranium(VI) and plutonium(IV) are extracted
during a multistage counter-current extraction process, leading to a
ca. 100~g/L load in U(VI). In the present study, we employed as organic
phase a 1~M TBP solution in $n$-dodecane, which was loaded with
uranium(VI) and thorium(IV) (as a substitute for plutonium(IV)) after a
one-stage liquid--liquid extraction of a synthetic 6~M HNO$_{3}$
aqueous phase, prepared from U(VI) and Th(IV) nitrate salts. The load
in U(VI) in the synthetic aqueous phase was close to saturation
(0.326~M  $\mathrm{UO}_{2}(\mathrm{NO}_{3})_{2}{\cdot}
6\mathrm{H}_{2}\mathrm{O}$), in order to approach the targeted
100 g/L U(VI) concentration in the organic phase. The Th(IV)
concentration in the synthetic aqueous phase (0.18~M
$\mathrm{Th}(\mathrm{NO}_{3})_{4}{\cdot}
5\mathrm{H}_{2}\mathrm{O})$ was chosen in order to reach a U:Th
ratio of about 9:1 in the organic phase after liquid--liquid
extraction, a value close to what is expected to occur (for Pu(IV))
when waste MOx fuel is recycled). Concentrations of U(VI) and Th(IV) in
the organic phase were therefore 0.28~mol/L and 33~mmol/L respectively,
i.e., 66 and 7.6~g respectively. 

From a general point of view, actinides(IV) and actinides(VI) are well
extracted into TBP-based organic solvents, contrary to actinides(III)
(and  actinides(V))~\cite{1}. As a consequence, during the PUREX
process, efficient stripping of plutonium is performed after reduction
of Pu(IV) into Pu(III), which is considered inextractible, with
distribution coefficients of about  10$^{-2}$. This step is required to
get complete decontamination of organic phase from plutonium. Use of a
low concentration HNO$_{3}$ aqueous solution leads to complete
stripping only when a high aqueous--to--organic phase volume ratio is
employed. This leads to complete stripping of U(VI) also, as exploited
for the quantification of both U and Th in organic phases employing an
aqueous--to--organic phase volume ratio equal to 100. 
Using a phase \mbox{volume} ratio equal to 1, with 
a 0.1~M HNO$_{3}$ aqueous solution,
stripping of Th(IV) is incomplete (ca.\ 90\%) and 35\% U(VI) is stripped
as well. Complete actinide(IV) stripping was therefore approached using
oxalic acid. As oxalic acid is not soluble in the organic phase, the
stripping of U and Th from the organic phase via precipitation was
performed by putting in contact the organic phase with an aqueous
C$_{2}$O$_{4}$H$_{2}$ solution. The latter was prepared by dissolving
oxalic acid dihydrate in a dilute HNO$_{3}$ solution or
in water. \looseness=-1

\begin{table*}
\caption{\label{tab1}Variation of the mass percentage of uranium and
thorium in the solid precipitate and organic phase as a function of the
concentration of C$_{2}$O$_{4}$H$_{2}$ after organic-phase
precipitation}
\begin{tabular}{ccccccccccccc} 
\thead
\xmorerows{1}{[C$_{2}$O$_{4}$H$_{2}$]} &
\multicolumn{2}{c}{0~M} & 
\multicolumn{2}{c}{0.057~M} & 
\multicolumn{2}{c}{0.085~M} & 
\multicolumn{2}{c}{0.114~M} & 
\multicolumn{2}{c}{0.142~M} &
\multicolumn{2}{c}{0.214~M} \\
\cline{2-3}\cline{4-5}\cline{6-7}\cline{8-9}\cline{10-11}\cline{12-13}
 & U & Th & U & Th & U & Th & U & Th & U & Th & U & Th \\ 
\endthead
Organic & 
65\% & 
9\% & 
68\% & N.D. & 63\% & N.D. & 56\% & N.D. & 51\% & N.D. & 37\% & N.D. \\ 
Solid &  &  & 0\% & 84\% & 0\% & 99\% & 13\% & 99\% & 23\% & 99\%  & 46\% & 99\% \\ 
(U:Th)$_{\mathrm{s}}$ &  &  & 
\multicolumn{2}{c}{0} & 
\multicolumn{2}{c}{0} & 
\multicolumn{2}{c}{1.1} & 
\multicolumn{2}{c}{1.9} & 
\multicolumn{2}{c}{3.9}
\botline 
\end{tabular}
\tabnote{Notes: [HNO$_{3}$] is fixed at 0.1~M, and the complement
to 100\% represents the fraction of the considered element in the
aqueous phase.}
\vspace*{-4pt}
\end{table*}

The coprecipitation of U and Th was carried out through dropwise
addition of the actinide solution into the aqueous
C$_{2}$O$_{4}$H$_{2}$ solution, using a volume ratio of the precursor
solutions equal to one. All experiments were conducted at room
temperature in a controlled environment 
$(20\pm2~\text{\textdegree}\mathrm{C})$, and temperature effect was not
studied. In a first set of experiments, the concentration of
C$_{2}$O$_{4}$H$_{2}$ in the initial aqueous solution was varied
between 0 and 0.214~M, while fixing that of HNO$_{3}$ at 0.1~M (in
anticipation of work with Pu(IV) which readily forms hydroxides unless
in very acidic media). In these experiments, the quantity of
C$_{2}$O$_{4}$H$_{2}$ was maintained lower than the quantity required
for the total precipitation of both actinides present in the organic
solvent (i.e., the sum of the molar amount of U(VI) and twice the molar
amount of Th(IV)), in order to observe selectivity between U and Th
precipitation. It was rapidly established that Th precipitates
preferentially, even in the presence of U. Based on analysis of both
aqueous and organic phases, at [C$_{2}$O$_{4}$H$_{2}$] ${<}$ 0.114~M,
thorium was totally precipitated  (Table~\ref{tab1}), whereas no traces
of uranium were detected in the precipitate. When the amount of
C$_{2}$O$_{4}$H$_{2}$ required for complete Th(IV) precipitation
(0.07~M expressed in concentration, twice the amount of Th(IV)) was
reached, U(VI) precipitation was not observed, and U(VI) was simply
stripped into the aqueous phase. At 0.085~M C$_{2}$O$_{4}$H$_{2}$
concentration, about 37\% of U(VI) were stripped from the organic
phase, with no U(VI) detected in the precipitate. Only at
[C$_{2}$O$_{4}$H$_{2}$] ${\geq}$ 0.114~M did uranium start
precipitating, with the precipitation yieldincreasing as expected with
the increase in C$_{2}$O$_{4}$H$_{2}$ concentration until it reached
46\% at 0.214~M. Th was always found in the precipitate (as long as
sufficient C$_{2}$O$_{4}$H$_{2}$ was added), but even at high
C$_{2}$O$_{4}$H$_{2}$ concentration, U(VI) was distributed between the
3 phases, with non-negligible amount in the aqueous phase. The change
in order of addition (i.e., addition of aqueous\unskip\break 
phase into organic
solution) did not change the outcome of the reaction. The U(VI)
precipitation was also found to have no kinetic limitation since rapid
precipitation was observed, with no evolution after 30~min. The latter
observation also revealed that it is difficult to decipher the detailed
mechanism of precipitation stripping, i.e., to make the difference
between direct precipitation in the organic phase after extraction of
C$_{2}$O$_{4}$H$_{2}$ into the organic phase, or stripping of metal
cations into the aqueous phase followed by precipitation with oxalic
acid in the aqueous phase. As oxalic acid is poorly soluble in organic
media, it can be suspected however that the precipitation occurs in the
aqueous phase, or at least at the interface between both phases.

PXRD analysis of the solids obtained (Figure~\ref{fig1}) revealed the
precipitation of thorium in the form of thorium oxalate hexahydrate
$(\mathrm{Th}(\mathrm{C}_{2}\mathrm{O}_{4})_{2}
{\cdot}6\mathrm{H}_{2}\mathrm{O})$ in all cases. At the concentration at
which uranium starts precipitating (i.e., when [C$_{2}$O$_{4}$H$_{2}$]
${\geq}$ 0.114~M), in addition to the peaks corresponding to 
$\mathrm{Th}(\mathrm{C}_{2}\mathrm{O}_{4})_{2}{\cdot}6\mathrm{H}_{2}\mathrm{O}$, 
peaks between 8
and  9\textdegree\ (2${\uptheta}$) started to appear, which surprisingly
do not correspond to uranyl oxalate trihydrate
$(\mathrm{UO}_{2}(\mathrm{C}_{2}\mathrm{O}_{4}){\cdot}
3\mathrm{H}_{2}\mathrm{O})$. It is also important to note
that the intensities of these peaks increased with increasing
concentrations of C$_{2}$O$_{4}$H$_{2}$. Furthermore, these peaks could
not be identified based on the ICDD (International Center for Crystal
Data) and, thus correspond to an unknown phase. The expected peaks for
uranyl oxalate were not detected in the diffractogram 
(Figure~\ref{fig1}).

\begin{figure*}
\vspace*{-1pt}
\includegraphics{fig01}
\vspace*{-2pt}
\caption{\label{fig1}PXRD patterns of the precipitates obtained from
organic-phase precipitation of uranium and thorium at variable
C$_{2}$O$_{4}$H$_{2}$ concentrations and a fixed HNO$_{3}$
concentration of  0.1~M. Reference diffractograms of $\mathrm{UO}_{2}
(\mathrm{C}_{2}\mathrm{O}_{4})
{\cdot}{3}\mathrm{H}_{2}\mathrm{O}$ and 
$\mathrm{Th}(\mathrm{C}_{2}\mathrm{O}_{4})_{2}{\cdot}
6\mathrm{H}_{2}\mathrm{O}$ correspond to JCPDS No.~01-073-7325
and No.~00-022-1485, respectively.}
\vspace*{-2pt}
\end{figure*}

With the aim of determining the nature of the elements constituting the
precipitates resulting from the organic-phase precipitation at
different C$_{2}$O$_{4}$H$_{2}$ concentrations, EDX analyses were
carried out. In addition to Th, U, C, and O, the spectra obtained
revealed the presence of phosphorus in the 
powder\unskip\break 
obtained for
[C$_{2}$O$_{4}$H$_{2}$] ${\geq}$ 0.114~M, whereas no phosphoros was
detected at lower concentrations for which only Th was precipitated.
These results may indicate that uranium, phosphorus and oxalate
precipitate simultaneously. Furthermore, as TBP is the only source of
phosphorus in the precipitation medium, it can be assumed that the
uranium solid obtained probably contains TBP. Solids composed of
uranium(VI) and analogs of TBP, including trimethylphosphate (TMP) and
triisobutylphosphate, have already been reported in the 
literature~\cite{15,16}. Furthermore, Rahman et~al.\ reported the
formation of a TBP-containing solid during the precipitation stripping
of U(VI) from TBP/kerosene by NH$_{4}$OH~\cite{10}. The presence of TBP
in the solid was confirmed through complementary experiments that
involved carrying out the precipitation of only uranium under the same
precipitation stripping conditions (in terms of C$_{2}$O$_{4}$H$_{2}$
and HNO$_{3}$ concentrations). The results obtained are presented in
the supporting information along with further characterization of the
uranium precipitate. These analyses proved the presence of oxalate,
nitrate or nitric acid, and uranium in the solid, but could not allow
determination of its precise composition. Thus, the mixed-composite
uranium solid will be referred to as U--TBP--Ox in the following. 

When a phosphorus-containing uranium compound is calcined, a uranium
phosphate is expected. Such a product is incompatible with the
objectives of this work, which aims at obtaining precursors of actinide
dioxides for MOx fuel. Uranium phosphates are minerals known for their
relatively high chemical stability~\cite{17}, once formed, and their
transformation into uranium oxide is not possible. It is therefore
essential to obtain only actinide oxalates at the end of the
precipitation from the organic phase, so as not to introduce any
phosphate impurities during the step of conversion into oxide by
calcination, which could be difficult to  remove~\cite{18}.

Neither the mode of addition (organic into aqueous or vice versa) nor
the stirring speed affected the nature of the precipitate obtained.
Therefore, attempts towards precipitating uranyl oxalate from the
organic phase without the precipitation of U--TBP--Ox were mainly
focused on varying the concentration of C$_{2}$O$_{4}$H$_{2}$ and
HNO$_{3}$ in the precipitation medium. To this end, a new set of
experiments was carried out by increasing the HNO$_{3}$ concentration
in the precipitation medium to 2~M (compared to 0.1~M in the previous
set of experiments), whereas the oxalic acid concentration was varied
between 0.08 and 0.24~M. The results obtained showed that thorium still
totally precipitates at all C$_{2}$O$_{4}$H$_{2}$ concentrations,
whereas uranium partially precipitates for [C$_{2}$O$_{4}$H$_{2}$]
${\geq}$ 0.2~M  (Table~\ref{tab2}). It is important to note that,
compared to the previous set of experiments (realized at a lower
HNO$_{3}$ concentration), the precipitation yield 
of\unskip\break  
uranium decreased,
probably as U(VI) is more stabilized in the organic phase. This can be
related to the fact that increasing aqueous HNO$_{3}$ concentration
leads to an increase in the distribution ratio of U(VI) between an
aqueous nitrate phase and a 30\% TBP in  $n$-dodecane organic phase.
Consequently, the maximum U(VI):Th(IV) mass ratio reached 1.8, compared
to 3.3 at lower HNO$_{3}$ concentrations. However, compared to the
previous set of experiments realized at [HNO$_{3}$] ${=}$ 0.1~M, in
addition to the formation of 
$\mathrm{Th}(\mathrm{C}_{2}\mathrm{O}_{4})_{2}{\cdot}
6\mathrm{H}_{2}\mathrm{O}$, XRD analyses revealed the formation of
$\mathrm{UO}_{2}(\mathrm{C}_{2}
\mathrm{O}_{4}){\cdot}3\mathrm{H}_{2}\mathrm{O}$ at [HNO$_{3}$] ${=}$ 2~M,
with the intensity of the peaks corresponding to U--TBP--Ox being
relatively low and decreasing with the increase in
C$_{2}$O$_{4}$H$_{2}$ concentration (Figure~\ref{fig2}). This indicates
that increasing the concentrations of both C$_{2}$O$_{4}$H$_{2}$ and
HNO$_{3}$ in the precipitation medium favors the precipitation of
uranyl oxalate compared to\unskip\break  
U--TBP--Ox. 

\begin{figure*}
\vspace*{3pt}
\includegraphics{fig02}
\vspace*{3pt}
\caption{\label{fig2}PXRD patterns of the precipitates obtained from
organic-phase precipitation of uranium and thorium at variable
C$_{2}$O$_{4}$H$_{2}$ concentrations and a fixed HNO$_{3}$
concentration of  2~M.}
\vspace*{6pt}
\end{figure*}

\begin{table*}
\caption{\label{tab2}Variation of the mass percentage of uranium and
thorium in the solid precipitate and organic phase as a function of the
concentration of C$_{2}$O$_{4}$H$_{2}$ after organic-phase
precipitation}
\begin{tabular}{ccccccccccc}
\thead
\xmorerows{1}{[C$_{2}$O$_{4}$H$_{2}$]} &
\multicolumn{2}{c}{0.08~M} & 
\multicolumn{2}{c}{0.12~M} & 
\multicolumn{2}{c}{0.16~M} & 
\multicolumn{2}{c}{0.2~M} & 
\multicolumn{2}{c}{0.24~M} \\
\cline{2-3}\cline{4-5}\cline{6-7}\cline{8-9}\cline{10-11}
 & U & Th & U & Th & U & Th & U & Th & U & Th \\ 
\endthead
Organic & 81\% & N.D. & 78\% & N.D. & 74\% & N.D. & 61\% & N.D. & 51\%
& N.D. \\ 
Solid & 0\% & 99\% & 0\% & 99\% & 5\% & 99\% & 24\% & 99\% & 33\% & 99\%  \\ 
(U:Th)$_{\mathrm{s}}$ & 
\multicolumn{2}{c}{0} & 
\multicolumn{2}{c}{0} & 
\multicolumn{2}{c}{0.4} &
\multicolumn{2}{c}{1.8} & 
\multicolumn{2}{c}{2.5} 
\botline 
\end{tabular}
\xxtabnote{Notes: [HNO$_{3}$] is fixed at 2~M, and the complement to
100\% represents the fraction of the considered element in the aqueous
phase.}
\end{table*}

Therefore, with the aim of precipitating uranyl oxalate while limiting
the precipitation of the\unskip\break  
U--TBP--Ox impurity, additional experiments were
carried out while further increasing both C$_{2}$O$_{4}$H$_{2}$ and
HNO$_{3}$ concentrations to 0.3 and 4~M, respectively. The XRD pattern
of the precipitate formed is shown in  Figure~\ref{fig3} (top). The
results obtained show the precipitation of both uranium and thorium in
the form of oxalates, with the complete absence of the U--TBP--Ox
impurity, which was also confirmed by EDX analysis, which revealed the
complete absence of phosphorus. Moreover, increasing the concentration
of C$_{2}$O$_{4}$H$_{2}$ and HNO$_{3}$ increased the\unskip\break 
precipitation
yield of uranium to 95\%, with Th still being totally precipitated and
the U--TBP--Ox impurity not detected.

\begin{figure*}
\includegraphics{fig03}
\vspace*{4pt}
\caption{\label{fig3}PXRD patterns of the precipitate obtained from
organic-phase precipitation of uranium and thorium under optimized
conditions ([C$_{2}$O$_{4}$H$_{2}$] ${=}$ 0.3~M and [HNO$_{3}$] ${=}$ 4~M)
after washing with dodecane (top) and washing with water (bottom).}
\vspace*{4pt}
\end{figure*}

The control of the U:Th mass ratio during the precipitation experiment
seems to be the result of a higher solubility of uranyl oxalate 
($\mathrm{LogKs} = {-}8.5$~\cite{19}) compared to thorium oxalate
($\mathrm{LogKs}= {-}24$~\cite{20}). As the latter compound is more
stable, it is reasonable to suppose that its precipitation occurs more
quickly. Furthermore, uranyl oxalate is expected to be slightly soluble
in water at  20~\textdegree C, with a\unskip\break 
solubility of about  23~mg/L
(0.05~mmol/L, corresponding to 13~mg/L uranium). On the contrary,
thorium oxalate has a very low solubility, in the nanomolar range.
Thus, a post-synthetic treatment of the precipitate was proposed in
order to remove uranyl oxalate from thorium oxalate through washing:
the precipitate was washed with distilled water. The elimination of
uranyl oxalate could be observed by the color of the precipitate which
changed from pale yellow to white after the second wash, as uranyl
oxalate is characterized by its pale yellow color. Moreover, ICP
analysis performedon the supernatant revealed the presence of uranium
(no traces of thorium) after the first and second wash. XRD analysis
(Figure~\ref{fig4}) performed on the precipitate revealed the presence
of only  $\mathrm{Th}(\mathrm{C}_{2}\mathrm{O}_{4})_{2}{\cdot}
6\mathrm{H}_{2}\mathrm{O}$ with no traces of
$\mathrm{UO}_{2}(\mathrm{C}_{2}\mathrm{O}_{4}){\cdot}
3\mathrm{H}_{2}\mathrm{O}$.

\begin{figure*}
\includegraphics{fig04}
\caption{\label{fig4}SEM analysis of the precipitate obtained from
organic-phase precipitation of uranium and thorium under optimized
conditions ([C$_{2}$O$_{4}$H$_{2}$] ${=}$ 0.3~M and [HNO$_{3}$] ${=}$ 4~M):
SE mode (left) and EDX-based elemental mapping of Th and U (right).}
\end{figure*}

Therefore, following the oxalate precipitation from the organic phase,
the uranium- and thorium-based precipitates can be separated through
washing with water, and the U:Th ratio in the solid material can be
further adjusted with Th always remaining in the solid. This allows a
perfect control of the oxalate mixture composition before calcination.
Uranium can be recovered from the remaining phases, and recovered
through precipitation after addition of further C$_{2}$O$_{4}$H$_{2}$.
The proposed approach thus enables rapid recovery of Th(IV) from the
U:Th mixture in the organic phase, either as such, or in the presence
of a controlled amount of U(VI). Such studies pave the way towards the
reprocessing of U(VI)/Pu(IV) phases, in order to enable multirecycling
of Pu(IV) into MOx fuel.

These results suggest that uranium and thorium precipitate separately
and do not coprecipitate to form a single material. Based on SEM
analyses, the precipitate obtained exhibits a platelet shape 
(Figure~\ref{fig4}). Moreover, EDX analysis reveals that U and Th are
not systematically colocalized: some microdomains are Th-rich, whereas
some others are U-rich  (Figure~\ref{fig4}). Altogether, whether the
precipitation occurs in organic or aqueous phase, or at the interface,
Th(IV) precipitates first as thorium oxalate, followed by U(VI) as
uranyl oxalate as soon as acid content of the aqueous phase is
sufficient. The U:Th ratio is monitored through either adjustment of
the C$_{2}$O$_{4}$H$_{2}$ concentration, or after precipitation via
selective leaching of uranyl oxalate.

\section{Conclusion}\label{sec4}
During this study, we established the possibility of recovering Th(IV)
by precipitation stripping using oxalic acid C$_{2}$O$_{4}$H$_{2}$, in
the presence of a controlled amount of U(VI), from an organic phase
mimicking that which arises from the classical PUREX process. Th(IV) is
quantitatively stripped as long as sufficient C$_{2}$O$_{4}$H$_{2}$ is
introduced, whereas U(VI) only partially precipitates unless excess
C$_{2}$O$_{4}$H$_{2}$ is employed. Furthermore, the results obtained
show that clean precipitation of U and Th from the organic phase in the
form of pure actinide oxalates was attained by controlling the
concentrations of C$_{2}$O$_{4}$H$_{2}$ and HNO$_{3}$ in the
precipitation medium. The U:Th ratio in the final solid is controlled
by the amount of C$_{2}$O$_{4}$H$_{2}$ employed, and can be further
adjusted through washing with water as
$\mathrm{UO}_{2}(\mathrm{C}_{2}\mathrm{O}_{4}){\cdot}
3\mathrm{H}_{2}\mathrm{O}$ is more soluble in water than
$\mathrm{Th}(\mathrm{C}_{2}\mathrm{O}_{4})_{2}{\cdot}
6\mathrm{H}_{2}\mathrm{O}$. Compared to the PUREX process currently
employed at industrial scale to recover U and Pu from waste nuclear
fuel for the further production of MOx, precipitation stripping seems
to be highly promising, as it allows the stripping of actinide(IV)
without the aid of a reducing agent. Furthermore, this work opens the
door for the application of precipitation stripping using other organic
ligands as precipitating agents, which may lead to various morphologies
of the uranium- and thorium-based precipitates and, eventually, to
better control over the calcination and pelletization steps. 

\section*{Declaration of interests}
The authors do not work for, advise, own shares in, or receive funds
from any organization that could benefit from this article, and have
declared no affiliations other than their research organizations.

\back{}

\section*{Supplementary data}
Supporting information for this article is available on the journal's
website under \printDOI\ or from the author. 

\CDRsupplementaryTwotypes{supplementary-material}{\cdrattach{crchim-320-suppl.pdf}}

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