Estela Reinoso-Maset
email: estela.reinosomaset@gmail.com

Andjacques Ly

Study of uranium(VI) and radium(II) sorption at trace level on kaolinite using a multi-site ion exchange model

Keywords: Uranium(VI), radium, kaolinite, sorption, ion exchange

Highlights

 sorption of trace U(VI) and Ra(II) on kaolinite was studied  multi-site sorbent/multi-species sorbate ion exchange model used to describe sorption  results provided U and Ra sorption equilibria stoichiometry and associated constants  Ra 2+ was the main adsorbed species over the pH range  U negatively charged hydroxylated forms dominated sorption at pH>8 Highlights 1

Introduction

Considerable large volumes of mill tailings are disposed every year near uranium mining sites [START_REF] Abdelouas | Uranium Mill Tailings: Geochemistry, Mineralogy, and Environmental Impact[END_REF]. Since most of the ore deposits have low uranium content, the mill tailings still comprise a small portion of the initial uranium and almost all of the daughter radionuclides of the 238 U decay series, including 226 Ra [START_REF] Déjeant | Field analyses of 238U and 226Ra in two uranium mill tailings piles from Niger using portable HPGe detector[END_REF][START_REF] Nirdosh | Radium in uranium mill tailings -Some observations on retention and removal[END_REF]. These decay products add to the environmental impactof uranium-waste, making the mill tailings to be considered low-level radioactive waste. In particular, 226 Ra is of notable relevance due to its long half-life (1600 y), degree of radiotoxicity and its short-lived Manuscript Click here to download Manuscript file: Reinoso-Maset & Ly_JEnvRad_2015_manuscript.docx Click here to view linked References decayproducts, such as gaseous 222 Rn. The environmental concern of these tailings (and any contaminated land) arises from the potentialreleaseof radionuclides into the groundwater, and, as a result, contamination of subsurface environments in the vicinity ofthe mine [START_REF] Phrommavanh | Characterizing the Transport of Natural Uranium and its Decay Product 226Ra, Downstream from Former Mines in France[END_REF].

As thoroughly reviewed in [START_REF] Payne | Guidelines for thermodynamic sorption modelling in the context of radioactive waste disposal[END_REF], a key factor on migration predictions is the extent of radionuclide sorption onto the surfaces of solid phases found in natural environments. These solid phases must be methodically characterised, with special emphasis on the identification of sorption sites, the estimation of their concentration and their sorption behaviour towards radionuclides under different conditions. The sorption of U(VI) and, in less extent, Rahave been previously studied for different iron oxyhydroxide, carbonate and clayey minerals [START_REF] Catalano | Uranyl adsorption onto montmorillonite: Evaluation of binding sites and carbonate complexation[END_REF][START_REF] Jones | Reactions of radium and barium with the surfaces of carbonate minerals[END_REF][START_REF] Payne | Surface complexation model of uranyl sorption on Georgia kaolinite[END_REF][START_REF] Sajih | Adsorption of radium and barium on goethite and ferrihydrite: A kinetic and surface complexation modelling study[END_REF][START_REF] Stammose | Sorption mechanisms of three actinides on a clayey mineral[END_REF][START_REF] Tachi | Experimental and modeling studies on sorption and diffusion of radium in bentonite[END_REF][START_REF] Turner | Surface-charge properties and UO 2 2+ adsorption of a subsurface smectite[END_REF]. Clay minerals are abundant in geological formations and soils, and have the distinctive property of retaining ions,with high affinity for cationic species.The sorption process on clays is mainly via ion exchange on the surface, becoming even more important with trace-level concentrations and low ionic strength conditions, as the onesin natural environments. [START_REF] Catalano | Uranyl adsorption onto montmorillonite: Evaluation of binding sites and carbonate complexation[END_REF]studied the sorption of U on montmorillonite and observed that the fraction of uranyl sorbed via cation exchange under low ionic strength conditions was more important than observed in previous studies, where only surface complexation was considered. [START_REF] Turner | Surface-charge properties and UO 2 2+ adsorption of a subsurface smectite[END_REF]used an ion exchange model to describe sorption of uranyl on a smectite type clay, and obtained equilibrium constants for the different sorption sites on the surface. An ion exchange model (considering two types of sorption sites) was also more suitable than surface complexation model to describe the retention of U(VI) at trace level on a clay mixture of kaolinite and smectite [START_REF] Stammose | Sorption mechanisms of three actinides on a clayey mineral[END_REF], and the sorption of Ra was also demonstrated to happen via an ion exchange mechanism on purified smectite [START_REF] Tachi | Experimental and modeling studies on sorption and diffusion of radium in bentonite[END_REF].

When radionuclides are found at trace or ultra-trace level, major cations (found at higher aqueous concentrations) are strong competitors for clay sorption sites in natural environments [START_REF] Tertre | Influence of the ionic strength and solid/solution ratio on Ca(II)-for-Na+ exchange on montmorillonite. Part 1: Chemical measurements, thermodynamic modeling and potential implications for trace elements geochemistry[END_REF]. Therefore, a full characterisationof the clay properties towards the major ions is also required to describe the sorption of trace metals on clays. This type of characterisation has already been carried out, for example, with argillaceous rock [START_REF] Motellier | Modelling of the ion-exchange properties and indirect determination of the interstitial water composition of an argillaceous rock. Application to the Callovo-Oxfordian low-water-content formation[END_REF],beidellite [START_REF] Robin | Ion exchange reactions of major inorganic cations (H + , Na + , Ca 2+ , Mg 2+ and K + ) on beidellite: Experimental results and new thermodynamic database. Toward a better prediction of contaminant mobility in natural environments[END_REF]and kaolinite [START_REF] Reinoso-Maset | Study of major ions sorption equilibria to characterize the ion exchange properties of kaolinite[END_REF]. In these three studies, the sorption of major ions was described using a multi-site ion exchange model, and provided with a comprehensive and robust sorption data setsthat can be applied when studying sorption of trace elements on these clays.

Therefore, the aim of this work was to study the sorption of trace level uranium(VI) and radium(II) on a common phyllosilicate mineral, kaolinite, in the presence of excess K, a common groundwater cation, in order to determine the ion-exchange equilibria occurring at the mineral-solution interface. This work has followed a multi-site ion exchange model that can describe chemical systems such as a multi-site sorbent/multispecies sorbate, and which has already been applied in Reinoso-Maset and Ly (2014) to successfully characterise the ion exchange properties of the same kaolinite used in this work. The model is presented in detail therein, thus here we focused on the part of the model that describes sorption of a trace element. Following a detailed experimental protocol using chemical and radiochemical analytical techniques, batch experiments over a wide pH range (from 2 to 11)were carried out to measure the distribution coefficient (K d ) of traceU(VI) and Ra(II) sorption on kaolinite.

Additionally, sorption isotherms of both elements were measured over a representative concentration rangeand at two different solution pH.Finally, to obtain fundamental sorption equilibria stoichiometry and associated constants, the sorption data for both elements was processed according to the ion exchange model described here.

Methodology

Reinoso-Maset and Ly (2014) obtained the type and concentration of sorption sites of the kaolinite used in this work, the related major ions sorption equilibria occurring at the mineralsolution interface and their associated selectivity coefficients (expressed as the corrected selectivity coefficients, K*). The main formalisms and assumptions of the model are presenttherein, hence we have focused this section on the part of the model describing sorption of an element at trace level in the presence of two competitor ions.

Adsorption on major sorption sites

The kaolinite crystal structure, Al 2 Si 2 O 5 (OH) 4 , presents permanent negatively charged sites, X i , and hydroxylated sites with pH-dependent charge, Y j OH, that can result in different sorption equilibria occurring at the solid solution interface. When the element M is found at trace level compared to a major competitor N + and in the absence of other ligands than OH -, the adsorption of M on a major sorption site can occur under its free M m+ form or as one of its hydroxylated species, M(OH) x m-x , depending on the solution pH.

Therefore, when consideringH + as the reference competitor ion, the sorption of freeor hydroxylated species of Mand N + on permanently charged sorption sites, X i , can be described by Equations 1 and 3 respectively. Their associated equilibrium constants (as the corrected selectivity coefficients) have the form of Equations 2 and 4.

M 𝑚 + + 𝑝X 𝑖 -H + + 𝑚 -𝑝 H 2 O ↔ X 𝑖 -𝑝 M OH 𝑚-𝑝 𝑝+ + 𝑚H + (1) 𝐾 M OH 𝑚 -𝑝 𝑚H * 𝑖 = X 𝑖 -𝑝 M OH 𝑚-𝑝 𝑝+ H + 𝑚 X 𝑖 -H + 𝑝 [M 𝑚 + ] • 𝑦 H 𝑚 𝑦 M (2) N + + X i -H + ↔ X i -N + + H + (3) 𝐾 N H * 𝑖 = X i -N + H + X i -H + [N + ] • 𝑦 H 𝑦 N (4)
where K* i is the corrected selectivity coefficient for M m+ or N + and H + exchange on a surface site;i is the index number of the surface site;m is the cation charge; X 𝑖 -𝑝 M OH 𝑚 -𝑝 𝑝+ , X i -N + and X i -H + are the chemical forms describing the sorption state of M, N and H + on site X i respectively; the square brackets [ ] indicate the concentration of species either in solution or adsorbed to site X i (in mol L -1 and mmol g -1 of dry solid respectively); and y H , y M and y N are the molar activity coefficients for H + , M m+ and N + quantified in the present work according to Davies expression, log 𝑦 𝑖 = -0.5

• z i 2 • 𝐼 1+ I -0.3 • 𝐼 , with ionic strength 𝐼 = 1 2 • C i z i 2 i
, where C i andz i are the molar concentration and charge of the species i respectively.

In the case of hydroxylated sorption sites, Y j OH, the sorption of anions (L -and anionic M hydroxylated species) must be also considered. Equations from5 to 10showall the chemical equilibria involved and their associated thermodynamic constants.

N + + Y 𝑗 OH ↔ Y 𝑗 O -N + + H + (5) 𝐾 N H * 𝑗 = Y 𝑗 O -N + H + Y 𝑗 OH [N + ] • 𝑦 H 𝑦 N (6) L -+ Y 𝑗 OH + H + ↔ Y 𝑗 OH 2 + L - (7) 𝐾 L * 𝑗 = Y 𝑗 OH 2 + L - Y 𝑗 OH [L -] H + • 1 𝑦 L 𝑦 H (8)
-if M(OH) x m-x are cationic species, 0 ≤ x = m -p ≤ m-1 and 1 ≤ p ≤ m, and thus:

M 𝑚 + + 𝑝Y 𝑗 OH + 𝑚 -𝑝 H 2 O ↔ Y 𝑗 O - 𝑝 M OH 𝑚-𝑝 𝑝+ + 𝑚H + (9)
-if M(OH) x m-x are anionic species, x > m and q ≥ 1, and thus:

M 𝑚 + + 𝑞Y 𝑗 OH + 𝑚H 2 O ↔ Y 𝑗 𝑞 M OH 𝑚+𝑞 + 𝑚H + (10)
Since the exchange reactions ofthe element M are in competition with the ones involving N + , as well as L -, the total concentration of sites X i and Y j OH(SC i and SC j respectively) will be given by Equation 11 and 12.

SC 𝑖 = X 𝑖 -H + + X 𝑖 -N + + 𝑝 X 𝑖 -𝑝 M OH 𝑚-𝑝 𝑝+ 1 𝑝=𝑚 (11) SC 𝑗 = Y 𝑗 OH + Y 𝑗 O -N + + Y 𝑗 OH 2 + L -+ 𝑝 Y 𝑗 O - 𝑝 M OH 𝑚-𝑝 𝑝+ 1 𝑝=𝑚 + 𝑞 Y 𝑗 𝑞 M OH 𝑚+𝑞 𝑞≥1 (12)
However, the concentration of sites occupied by the trace metal M (either in Equation 11or 12) can be considered negligible respect to the concentration of sites with adsorbed H + , N + and L -, and thus the distribution coefficient of the element M (K d M = M M ) canbe expressed as a function of the corrected selectivity coefficients of M, N and L, the concentration of sites i and j(SCi and SCj) and the solution pH by combining Equations 2, 4, 6, 8 11 and 12. The final mathematical development yields to Equation 13.

K d M = 1 α M • H + 𝑚 • 𝑦 H 𝑚 𝑦 M • 𝐾 M OH 𝑚 -𝑝 𝑚H * 𝑖 • SC 𝑖 𝑝 1 + 𝐾 N H * 𝑖 • N + H + • 𝑦 N 𝑦 H 𝑝 1 𝑝=𝑚 sites 𝑖 + 𝐾 M(OH ) 𝑚 -𝑝 /𝑚H * 𝑗 • SC 𝑗 𝑝 1 + 𝐾 N/H * 𝑗 • N + H + • 𝑦 N 𝑦 H + 𝐾 L * 𝑗 • L -• H + • 𝑦 L • 𝑦 H 𝑝 1 𝑝=𝑚 sites 𝑗 + 𝐾 M(OH ) 𝑚 +𝑞 /𝑚H * 𝑗 • SC 𝑗 𝑞 1 + 𝐾 N/H * 𝑗 • N + H + • 𝑦 N 𝑦 H + 𝐾 L * 𝑗 • L -• H + • 𝑦 L • 𝑦 H 𝑞 𝑞≥1 (13)
where α M is the overall side-reaction coefficient expressed as shown in the Supplementary Information, Table S1.

𝛼 M = [M] total [M 𝑚 + ] = 1 + 𝛽 𝑟 H + 𝑟 + 𝛽 𝑠 • 𝑠 L 𝑙-𝑠
If the terms in Equation 13 that are independent to i and j are arranged with the K d , an equation with the form of a Y = F(X, Z) function can be obtained (Equation 14):

𝐘 = log 𝐾 M OH 𝑚 -𝑝 𝑚H * 𝑖 • SC 𝑖 𝑝 1 + 𝐾 N H * 𝑖 • 10 𝐗 𝑝 1 𝑝=𝑚 sites 𝑖 + 𝐾 M(OH ) 𝑚 -𝑝 /𝑚H * 𝑗 • SC 𝑗 𝑝 1 + 𝐾 N/H * 𝑗 • 10 𝐗 + 𝐾 𝐿 * 𝑗 • 10 𝐙 𝑝 1 𝑝=𝑚 sites j + 𝐾 M(OH ) 𝑚 +𝑞 /𝑚H * 𝑗 • SC 𝑗 𝑞 1 + 𝐾 N/H * 𝑗 • 10 𝐗 + 𝐾 𝐿 * 𝑗 • 10 𝐙 𝑞 𝑞≥1 (14)
where Y, X and Zcontain factors that can be obtained experimentally or calculated .

𝐘 = log K d M • α M • H + m • (𝑦 H ) m 𝑦 𝑀 (15) 𝐗 = log N + H + • 𝑦 N 𝑦 H (16) 𝐙 = log L -• H + • 𝑦 L • 𝑦 H (17)
Equation 14 is a logarithm of a sum of different terms describing the following possible sorption processes occurring in the system:

-cation adsorption on permanently charged sites X i

Y i(p) = log K M OH m-p /H *i • SC i p 1+K N/H *i •10 X p = 𝑓 𝑖(𝑝) 𝐗 -cation adsorption on hydroxylated sites Y j OH Y j(p) = log K M OH m-p /H *j • SC j p 1+K N/H *j •10 X + K L *j •10 Z p = 𝑓 𝑖(𝑝) 𝐗, 𝐙 -anion adsorption on hydroxylated sites Y j OH Y j(q) = log K M OH m+p /OH *j • SC j q 1+K N/H *j •10 X + K L *j •10 Z q = 𝑓 𝑗 (𝑞) 𝐗, 𝐙 Therefore, when X - , then Y i(p)  log(K M OH m-p /H *i • SC i p ), which results in a horizontal asymptotefor the Y i(p) = f i(p) (X) curve that allows estimating the value of the corrected selectivity coefficient for M (K *i ). WhenX+, then Y i(p)  log K M OH m-p /H *i • SC i p K N/H *i p
p• X, which isa linear asymptote with negative slopeindependently of the value of p, always a positive number (1 ≤ p ≤ m). This slope provides the stoichiometry of the adsorbed species and the associated chemical equilibrium.

In the same way, andif K N/H *j .10

X ≫ K L *j .10 Z ,Y j(p)  log K M OH m-p /H *j • SC j p K N/H *j p -p• X and Y j(q)
 log

K M OH m+p /OH *j • SC j q K N/H *j q
q• X when X +, defininglinear asymptotes with negative slopes (1 ≤ p ≤ m andq ≥ 1). Since X and Z vary in the opposite direction with increasing pH (see Equations 16 and 17), the increase of Y with X (i.e. positive slope of the curve Y = f(X))

implies that K N/H *j .10 X ≪K L *j .10 Z , which gives evidence of hydroxylated sorption sites that are able to absorb anionic species of M.

The curvesY = f(X) and Y = g(Z)can be obtained experimentally by determining the distribution coefficient of M in the presence of excess N + over a wide pH range. The commonly used K d vs. pH curve can then be converted to Y = f(X) and Y = g(Z)curves representation, where the charge of the adsorbed species p and q can be determined through the slopes of the curves, which consequently allows to write the stoichiometry of the sorption reaction for the trace element M (Equations 1, 9or 10). The corrected selectivity coefficient associated to the reaction on each sorption site,

K M OH m-p /H *i , K M OH m-p /H *j and K M OH m+p /OH *j
, can be obtained by nonlinear fitting of the experimental data. The values of SC i , SC j and 𝐾 N/H * 𝑖 , 𝐾 N/H * 𝑗 , K L *j need to be determined a priori from saturation curves of major cations and anions over pH. The clay material used in Reinoso-Maset and Ly ( 2014)is the same as the one used in this work, therefore site concentrations and stability constants obtained therein were used in the calculations presented here (see discussion).

Adsorption on minor sorption sites

Minor sorption sites (<0.001 mmol g -1 of dry clay) can have a significant affinity for elements at trace level.Reinoso-Maset and Ly(2014) identified one minor sorption site on the kaolinite after performing adsorption isotherms of Cs over a concentration range, at constant pH and excess of Na + in solution (i.e. a competitor ion for the sorption sites). Therein, the methodology and equations for the sorption of monovalent cations is fully explained. In this work, the trace elements are present in solution mainly as their doubly charged species.The equation to calculate the total adsorbed concentration, M T , of a doubly charged elementM in the presence of a monovalent competitor cation can be derived in the same manner and has the final form of Equation 18.

M T = (X i -) 2 M 2+ n 𝑖=1 = = SC 𝑖 2 + 1 + 𝐾 N/H * 𝑖 . N + H + . 𝑦 N 𝑦 H 2 8 • K M/H *i . M 2+ H + 2 • y M y H 2 n 𝑖=1 - 1 + 𝐾 N/H * 𝑖 . N + H + . 𝑦 N 𝑦 H 2 8 • 𝐾 M/H * 𝑖 . M 2+ H + 2 • 𝑦 M 𝑦 H 2 2 + SC 𝑖 • 1 + 𝐾 N/H * 𝑖 . N + H + . 𝑦 N 𝑦 H 2 8 • 𝐾 M/H * 𝑖 . M 2+ H + 2 • 𝑦 M 𝑦 H 2 (18)
This equation can then be used, along with the measured M concentration in solution at (2014). By performing a nonlinear fitting of the adsorption isotherms (logK d vs. log M T ), the corrected selectivity coefficients associated to the adsorption equilibria reactions on minor sorption sites can be calculated.

Materials and methods

Reagents and solutions

All solutions were prepared with ultrapure water (18.2 MΩ•cm -1 , Milli-Q gradient system, Millipore) and analytical grade reagentspurchased from Sigma Aldrich (USA), Fisher Scientific (UK), Fluka (Germany) or VWR (France).Solid phase masses and solution volumes were always determined by weight using daily calibrated 4 decimal figure balances (Mettler Toledo AT200 or XP504, France). Experiments were carried out at room temperature and atmospheric CO 2 , except for the preparation of K hydroxide solutions (used to adjust the pH of the clay suspensions) which was carried out in CO 2 -free conditions to minimise additional complexation processes.

Radioisotopes solutions of 232 U and 226 Ra were obtained from Isotope Products Laboratories (USA). Due to the relatively short half-life of 232 U (t 1/2 = 68.9 years), the formation of decay products (i.e. 228 Th, 224 Ra, 220 Rn, etc.) can cause interferences in the activity counting, therefore the source solution was firstpurifiedfollowing the methodology proposed by [START_REF] Kraus | Anion-exchange studies. XXI. Th(IV) and U(IV) in hydrochloric acid. Separation of thorium, protactinium and uranium[END_REF]. Analiquot of the 232 U source solution wasloaded in an ion exchange resin column (AG1X8, 100-200 mesh, chloride form; Biorad Laboratories, USA)pre-conditioned with 8 M HCl. The decay products were washed out using 8 M HCl, and to recover the more strongly retained 232 U, the column was rinsed with 0.1 M HCl. The eluted solution was neutralised with a small volume of 4 M KOH. This fresh 232 U spiking solution was used within 21 days to ensure that less than 2% of the total activity was due to the accumulated decay products. In the case of 226 Ra (t 1/2 = 1600 years), the source and spiking solutions were aerated under the fume hood before sampling to avoid any external contamination by the daughter gas 222 Rn (t 1/2 = 3.825 days). The formation of other decay products was not an issue for the type of experiments carried out in this study.

Clay preparation

The kaolinite clay mineral used in the workwas 100% wt. pure kaolinite purchased from Sigma-Aldrich (Germany). Its purity was verified by X-ray diffraction and its N 2 -BET specific surface area was 21.38 m 2 g -1 (Reinoso-Maset and Ly, 2014). Prior to the batch experiments, the raw pure kaolinite was conditioned to its K homoionic form as explained in detail in (Reinoso-Maset and Ly, 2014).Briefly, raw kaolinite was first homoionized to Nakaolinite and then converted to K-kaolinite. A MilliQ water suspension was carried out to eliminate any residual salt solution, and a final ethanol suspension was used to enhance the drying process. The homoionic K-kaolinite was then gently ground to fine powder and stored in a constant relative humidity container until use. The water content calculated after drying few mg at 105°C was less than 1.30% and considered for the actual mass involved in each individual batch.

Adsorption over pH experiments

The adsorption of U(VI) and Ra(II) at trace level was studied at room temperature for a pH range between 2 and 12in the presence of excess K. The U and Ra distribution coefficients weredeterminedusing 0.1 g of K-kaolinite and 20 mL solution phase in tightly capped 30 mL Nalgene centrifuge tubes.This experimental configuration offered a minimal headspace and when sampling was needed the exposure to the laboratory air was for a short time, thus limiting the introduction of atmospheric CO 2 to a negligible amount. In addition, other potential sources of air CO 2 were prevented, e.g. KOH was prepared in a glove-box and added to clay suspensions via an automatic burette.

In order too fully disperse and hydrate the clay material, itwas first suspended with MilliQ waterand after > 16 h aliquots of KClO 4 solution were added accordinglyto obtain5•10 - 2 M K.The potassium perchlorate saltwas used to avoid any potential complexation of U or Ra by the salt anion.After > 16 h equilibration time, the pH was adjusted with known aliquots of concentrate HClO 4 and KOH (previously determined by titration of an equivalent clay amount)and suspensions were shaken for > 20 h, a sufficient time to reach thermodynamic equilibrium between solid and solution phases. The final solution pH was measured after phase separation by ultracentrifugation (20 min, 20000 rpm), and two 1 mL and one 5 mL supernatant aliquots were collected for anion, cation and dissolved Si analysis.

At this stage, an aliquot of known 232 U and 226 Ra activitywas added to each batch suspensions, corresponding to initial U and Ra concentrations of ca. 10 -9 and 10 -10 M respectively. In both cases, the spiked suspensions were equilibrated overnight on a TURBULA orbital mixer before solid and solution phases were separated by ultracentrifugation. A 1 mLaliquot of the supernatant was taken from uranium batch experiments for liquid scintillation counting (LSC) and 10 mL aliquot from Ra batch experiments forgamma spectrometry. Based on the isotopic dilution principle of nofractionation between stable and radioactive M, determining thesupernatant activityis a simple approach to measure the distribution ratio between solid and solution phases, and consequently the concentration of adsorbed U or Ra (Reinoso-Maset and Ly, 2014).

Adsorption isotherm experiments

The U(VI) and Ra(II)adsorption isotherms were obtained at two given pH using 0.15 g of K-kaolinite. Buffer solutions of2.5•10 -3 M KOH and 5•10 -3 M 2-[N-morpholine]ethanesulfonic acid (MES) or 3-[cyclohexylamino]-1-propane-sulfonic-acid (CAPS) were used to fix the solution pH at 6.1 and 10.2 respectively. These reagents are known to have no interaction with metallic elements [START_REF] Yu | Avoiding Interferences from Good's Buffers: A Contiguous Series of Noncomplexing Tertiary Amine Buffers Covering the Entire Range of pH 3-11[END_REF] and their anionic forms are unlikely to be adsorbed onto the clay surface, therefore, they will not interfere with the ion exchange reactions.The buffer solutions also contained 5•10 -2 M KClO 4 to provide excess K in solution.

In order to avoid pH changes during the adsorption process, the K-kaolinite was first washed with 20 mL of buffer solutions. The suspensions were mixed during 1 h and after ultracentrifugation (20 min, 20000 rpm) the supernatant was carefully removed to avoid clay loss. The buffer solution wash was repeated one more time. A final 20 mL buffer solution was used to equilibrate the system overnight. Clay dissolution and K concentration were checked after ultracentrifugation (20 min, 20000 rpm) by sampling 5 mL of the supernatant for Sianalysis, and two 0.1 mL aliquots for cation and anion determination.

In the U batch experiments, and without exceeding the solubility limit of U at each pH value, aliquots of stable U stock solutions prepared from an ICP-AES standard solution (1000 μg mL -1 U in 4% HNO 3 , PlasmaCAL, SCP Science, Canada) were added to the suspensions to obtain 10 concentrations between 2•10 -10 and 1•10 -7 M for the MES series (pH 6.1), and 12 concentrations up to 1.26•10 -6 M for the CAPS series (pH 10.4). The suspensions were placed on an orbital shaker and left to equilibrate for > 20 h. Both series were finally spiked with 200 μL of 232 U spiking solution prepared as described above. The additionof 232 U was taken into account in the total U concentration in solution.Stable Ra salts or solutions are not available, thusthe concentration range in the Ra batch experiments was obtained from a 226 Ra spiking solution. The added activities corresponded to an initial Ra concentration between 1.82•10 -10 and 2.73•10 -7 M for the 10 MES batch experiments (pH 6.1), and between1.12•10 -10 and 3.47•10 -7 M for the 10 CAPS batch experiments (pH 10.4). In both set of experiments, the spiked suspensions were shaken overnight on a TURBULA mixer before solution and solid phases were separated by ultracentrifugation (20 min, 20000 rpm) and 1 and 10 mL aliquots were taken for 232 U LSCand 226 Ra gamma spectrometry respectively.

Analytical methodology and experimental error

The solution pH, cation and anion concentration by ion chromatography and dissolved silicon concentration by UV-Vis spectrometry were determined in the exact same manner as the methodology followed in Reinoso-Masetand Ly (2014) and analytical details are explained therein.

The activity of 232 U was measured on a Tri-CARB 2700TR liquid scintillation analyser (Packard, USA) using the α/β discrimination mode to verifythat the activity is originated by a pure α emitter. Since the batch series contained high concentration of salt buffer solutions, 0.3 mL supernatant aliquots were diluted with 0.7 mL of acetate buffer (0.3 M final concentration) before mixing with 4 mL scintillation cocktail. This resulted in an equal quenching level for all samples with a constant counting efficiency. The counting was performed during 15 min and only the α emission band (200 -400 keV) was considered.

The γ activity of 10 mL supernatant aliquots containing 226 Ra was determined using a gamma spectrometer equipped with an N-type coaxial high purity germanium detector (ITECH-Instruments). The gamma spectra were collected from 0 to 2000 keV during 60 min and treated with the InterWinner 6.0 software. The 226 Ra band was identified at 186 keV. This band coincides with the 235 U energy band, but since U was not present in these batch solutions, any registered activity at this specific energy was attributed to the Ra isotope.

The uncertainties reported in this work were calculated for each experimental batch by the propagation of known measurement errors theory,as explained in detail inReinoso-Masetand Ly(2014).

Results and discussion

Adsorption of U and Ra over pH

The K d distribution for trace U and Ra sorption on K-kaolinite was obtained over pH in the presence of excess K (Figure 1). The U data shows a sorption edge at around pH 3.5 followed by a wide sorption plateau, that ends with a slight K d increase and another two short sorption edges are seen at the highest pH range of the experiments (pH 10-11). Radium however was completely adsorbed at the highest pH, causing null K d results and thus the data points above pH 10.6were excluded from the curve representation. Negligible sorption was observed below pH 4, resulting in K d values around 0. This data is shown in the graph for reference, but it was not considered in the data treatment. Between pH 4 and 10, the K d increased gradually with increasing pH, as expected from the exchange reaction. However, these results in this form do not provide further information about the charge of the adsorbed species nor the stoichiometry of the reaction.Therefore, the experimental K d data was transformed to the Y = f(X) and Y = g(Z) functions and Y was represented against X for both trace U and Ra sorption (Figure 1).

In the case of U, a small plateau is observed at 1 < X < 2.5, followed by a decrease of the curve with slope between -2 and -1, for then, at X > 6, increase with +1 and +2 slope up to X = 10. The positive slope observed in the Y = f(X) could be due to the precipitation of uranyl silicate minerals formed with Si brought into solution by the inevitable dissolution of kaolinite. However, special care was taken when choosing the U concentration for these experiments, by working with considerable low total concentration of the radioisotope. The non-precipitation was confirmed by the total U in solution at equilibrium for each experimental batch falling below the calculated limiting solubility curves for three U mineral phases at the experimental conditions (Figure S1in SI). Another consideration is the formation of uranyl carbonate species. [START_REF] Payne | Surface complexation model of uranyl sorption on Georgia kaolinite[END_REF] studied the sorption of U over the pH on two Georgia kaolinites and observed that at pH > 8 the complexation of U by dissolved carbonate reduced the U uptake. It was elsewhere mentioned that uranyl carbonate species could be adsorbed on another clay mineral surface, montmorillonite [START_REF] Catalano | Uranyl adsorption onto montmorillonite: Evaluation of binding sites and carbonate complexation[END_REF].

Both studies, however, were conducted on air-equilibrated clay suspensions, i.e. systems containing more than about 10 -5 M total dissolved carbonate. According to thermodynamic predictions, this dissolved concentration will exponentially increase at pH higher than 5.5

(Figure S2 in SI) and would explain the observed uranium reactivity. As mentioned above, in this work the amount of carbonate involved in each experiment was sufficiently small to prevent a significant formation of uranyl carbonate complexes.Therefore, the positive slopes were assigned to the sorption of negative uranyl species according with the dissolved species distribution in the absence of carbonate, which shows thatthe main speciesat pH > 8 are

UO 2 (OH) 3 -and UO 2 (OH) 4 2- (Figure S3 in SI).
The Y = f(X) representation for Ra shows a curve with two linear parts (1 < X < 7 and 7 < X < 9) both with slope -2, corresponding to a positively doubly charged species being adsorbed on two different sorption sites. This speciation of Ra agrees with previous studies (e.g. [START_REF] Sajih | Adsorption of radium and barium on goethite and ferrihydrite: A kinetic and surface complexation modelling study[END_REF]) and with its solution speciation distribution of Ra, which is dominated by Ra 2+ up to pH 11.5 in the absence of carbonate or any other complexing agent (Figure S3 in SI).

Once the charge of the adsorbed specieswere deduced, a nonlinear fitting of the Y = F(X, Z) representation for both U and Ra adsorption was performed using Equation 14. The number and concentration ofsorption sites on the kaolinite, as well as the constants describing the ion exchange reaction for the major cation and anion (K and ClO 4 in these experiments)

were obtained inour previous work (Reinoso-Maset and Ly, 2014). Briefly, cation and anion saturation curves revealed five major sorption sites on the kaolinite surface, of which two were identified as permanently charged sites (X 3 -, X 4 -) and three as hydroxylated sorption sites (Y 1 OH, Y 2 OH, Y 5 OH). The sorption isotherms confirmed the presence of a minor sorption site of lower concentration but higher affinity. Table S2 in SI shows all the solution species, sorption sites and equilibrium constants used to resolve Equation 14 for each system.After some trial and error fittings with all the possibilities, the calculated curves were simplified to the sum of the curves corresponding only to the sorption equilibria that contributed most at each pH interval (e.g. the sorption of negative uranyl species at low pH was negligible compared to the positively charged species). The Y = F(X, Z) fittings for U and Ra were optimized by nonlinear regression using Excel Solver tool and are represented in Figure 2.The stability constants, as corrected selectivity coefficients, obtained from the optimized fittings are summarised in Table 1. These corrected selectivity coefficients for U and Ra, along with constants previously obtained for K, ClO 4 and sorption site concentrations (Reinoso-Maset and Ly, 2014)were subsequently used in Equation 13to fit the logK d vs. pH graphs (Figure 1 for raw data). Figure 2 shows the individual curves for each U and Ra species as well as the overall calculated K d curve.

For U, the sorption of the free uranyl (UO 2 2+ ) was dominant at pH < 5 on sorption site Y 2 OH but also presented considerable sorption on X 4 at circumneutralpH. The positively charged hydroxylated form, UO 2 (OH) + , adsorbed mainly on site X 3 . The negatively charged species, UO 2 (OH) 3 -and UO 2 (OH) 4 2-

, were the predominant species adsorbed on site Y 5 OH at pH > 7. This distribution of the adsorbed species over pH highly agrees with the speciation in solution (Figure S2 in SI), and the overall calculated K d curve fits satisfactorilywell to the experimental data (see comparison of calculated vs experimental data in Figure 1).In the case of Ra, the lower number of experimental points limited the fitting to only few possible modelling assumptions. Despite higher uncertainty (see calculated vs experimental in Figure 1), the fitting reveals that the sorption of Ra 2+ over pH occurred more importantly on sites Y 2 OH and X 4 , behaviour that resembles the sorption of divalent cations, Ca and Mg, on the same kaolinite material(Reinoso-Maset and Ly, 2014).

Adsorption isotherms of U and Ra

Figure 3 shows the experimental adsorption isotherms of U and Ra at trace level at pH 6.1 and 10.4. Regardless the variability on the pH 10.4 series for U, a significant difference on the coefficient distribution was observed between the different pH series. This difference was more remarkable between the Ra series, demonstrating that as the pH increases higher sorption is observed, as seen in the saturation curves.

For both elements and both pH series, a decrease of sorption at around log[M] = -4.5 mmol g -1 proves the existence of a low concentration site, which was saturated as the concentration of U or Ra increased in solution. Since all series showed the same inflexion point, the concentration of this site is hence independent of the solution pH and solution speciation. Moreover, this estimated concentration (0.082 µmol g -1 )agrees with the minor site (X s ) identified on the Cs isotherm in Reinoso- [START_REF] Reinoso-Maset | Study of major ions sorption equilibria to characterize the ion exchange properties of kaolinite[END_REF]. Therefore, at the time of calculating the U and Ra isotherm curves by non-linear fitting, this value was used to determine the contribution of a minor site on the overall U and Ra sorption. In both cases, the contribution of a major site (i.e. Y 1 OH) was also taken in account, along with the corrected selectivity coefficients for K that had been obtained in previous experiments(Reinoso-Maset and Ly, 2014).

The calculated isotherms curves are shown in Figure 3. For trace U at both pH and Ra at pH 10.4, the major contribution to the sorption was by the minor site, X s , while for Ra at pH 6.1 the major site, Y 1 OH, had a significant contribution to the sorption. From the non-linear fitting, the best estimated of the corrected selectivity coefficients for U and Ra sorption on the minor site X S were calculated as well as the U sorption on site Y 1 OH, which had not been revealed in the saturation curve (Table 2). The comparison of calculated and experimental K d values shows a satisfactory agreement for both elements (Figure 3).

Conclusions

The treatment of common experimental K d values using the multi-site ion exchange model presented here provided with extra information about the speciation of sorbed U and Ra at trace level on kaolinite over a wide pH range. Moreover, the thermodynamic constants of the sorption equilibria were also obtained during optimisation of the model. Table 3 compiles the chemical equilibria for U and Ra and their associate equilibrium constants (as corrected selectivity coefficients). Ra was sorbed as its free form, Ra 2+ , from pH 2 to 10.5 on two main sorption sites, comparably to the behaviour of Ca and Mg sorption on kaolinite.

Uranium, on the other hand, showed a more complicated speciation with positively charged species (free and hydroxylated forms) dominating the sorption at pH below 7, while negatively charged uranyl species were sorbed at higher pHs on the most basic sorption site.

The sorption isotherms proved the existence of minor sites, which have lower concentration but high affinity for elements at trace level concentrations.

The sorption equilibria and associated constants reported here complete the major ions sorption equilibria obtained previously (Reinoso-Maset and Ly, 2014) and consolidate the multi-site ion exchange model as a reliable approach to obtain a robust thermodynamic databasedescribing the sorption of trace elements in the presence of major cations on kaolinite, and in any other similar multi-site ion exchanger. Consequently, this fundamental thermodynamic database will improve the knowledge on radionuclide behaviour in environments containing clayey minerals, which are extremelyrelevant for migration predictions at contaminate sites. Tables Table 1. Type and concentration of major sorption sites, charge of the adsorbed species and the associated corrected selectivity coefficients, 𝐾 [𝑴 𝑂𝐻 𝑥 ] 𝑚 -𝑥 (𝑚 -𝑥)H * 𝑖 𝐾 [𝑴 𝑂𝐻 𝑥 ] 𝑚 -𝑥 (𝑚 -𝑥)OH * 𝑗 describing the sorption of trace U(VI) and Ra(II) on K-kaolinite in the presence of 5•10 -2 M K. The values of the constants were determined after non-linear fitting of the Y = F(X, Z) representation of the U and Ra sorption (Equation 14). *values previously obtained [START_REF] Reinoso-Maset | Study of major ions sorption equilibria to characterize the ion exchange properties of kaolinite[END_REF] Table 2.Type and concentration of sorption sites and their associated corrected selectivity coefficients (K *i M/mH ) describing the sorption of trace U(VI) and Ra(II) on K-kaolinite in the presence of excess K. Values were determined after non-linear fitting of the U and Ra isotherms at pH 10.4 and 6.1 and [K] = 5•10 -2 M using Equation 18.

Figures

𝑟

  toaccount for the aqueous chemical equilibria, including the complexation of M by hydroxide ion(effective formation constant β r ) and ligands L l-(effective formation constantsβ s ).The reactions and constants to calculate the side-reaction coefficient for U and Ra in this work are

  equilibrium, M , to obtain the distribution coefficient (K d ) of the element M. Ata given pH and constant N + concentration, and providing thatM is at trace level compared to the concentration of sites, M T changes proportionally to M and thus the K d tends to a constant value. When M T reaches the concentration of minor sorption sites, SC s , a pronounced decrease of K d towards 0 occurs due to full saturation of the sorption sites. In Equation18, the solution pH and concentrations can be obtained experimentally, whereas the values of SC i and 𝐾 N/H * 𝑖 were determined from the cation saturation curves over pH in Reinoso-Maset and Ly

Figure 1 .Figure 2 .

 12 Figure 1. Left:Experimental distribution coefficient (K d ) over pHfor U (top) and Ra (bottom) sorption on K-kaolinite in the presence of 5•10 -2 M K.Grey triangles were excluded for further data analysis.Right: Y = f(X) representation of the same U and Ra sorption data following Equation14. Grey lines are 1:1 grids to help visualize the slopes. In all cases, the error bars represent ± 1 σ.

Figure 3 .

 3 Figure3. Adsorption isotherms of U (top) and Ra (bottom) on K-kaolinite. The distribution coefficient is represented as a function of adsorbed U or Ra concentration at pH 6.1 (empty symbols) and 10.4 (filled symbols), in the presence of 4.75•10 -2 M K. The dotted and dashed lines correspond to the calculated adsorption isotherms on a minor sorption site, Xs, and a major sorption sites, Y 1 OH; and the full red lines are the overall U and Ra adsorption isotherms fittings. Graphs on the right are the comparisonof experimental and calculated distribution coefficients. In all cases, the error bars represent ± 1 σ.

Tables Click here to download Table: Reinoso-Maset & Ly_JEnvRad_2015_tables.docxTable 3 .

 3 Stoichiometry and associated corrected selectivity coefficientsof trace level U(VI) and Ra(II) adsorption equilibria on kaolinite.

	Sorption	Concentration	Corrected selectivity coefficients
	Sites	(mmol g -1 )	𝑲 𝐔𝐎𝟐 𝐇 * 𝒊	𝑲 𝐑𝐚 𝐇 * 𝒊	𝑲 𝐊 𝐇 * 𝒊
	X S	0.000082 ± 0.000002*	35.4 ± 0.2 (pH 6) 50.3 ± 0.2 (pH 10)	0.121 ± 0.004 (pH 6) 17.5 ± 0.2 (pH 10)	(12.7±0.5)•10 -5
	Y 1 OH	0.0108 ± 0.0001*	3.56•10 -5	1292 ± 20*	(1.83 ± 0.05)*
	*values takenfrom Table 1and Reinoso-Maset and Ly (2014)

Acknowledgements

This research was partially funded by AREVA Mines Direction R&D under Contract number A-NOPRA-02-03.


L Bonato

M Virot

T Dumas

P Moisy

Sergey I Nikitenko

Ultrasound-assisted Conversion of ThO_2 into a Th peroxo sulfate

des établissements d'enseignement et de recherche français ou étrangers, des laboratoires publics ou privés.

Institut de Chimie Séparative de Marcoule (ICSM) -UMR5257, CNRS/CEA/UM/ENSCM, Site de Marcoule, 30207 Bagnols sur Cèze, France 2 CEA/DEN/MAR/DMRC, Nuclear Energy Division, Radiochemitry and Process Department, 30207 Bagnols sur Cèze, France

 oral  poster Nanostructured materials can be defined as materials with a microstructure the characteristic length scale of which is on the order of a few (typically 1-10) nanometers. [1] Nanostructuring of materials is an important subject of study because of their interesting physical and chemical properties which can be optimized according to their size, morphology and structure. The nanoscale particles that compose these materials have a large surface-to-volume ratio and provide a greater amount of active sites compared to other larger materials. Therefore, nanostructured materials may find applications in catalysis, biomedical, synthesis of luminescent materials, etc. Nevertheless, nanostructuring of actinide oxides and their subsequent reactivity are poorly reported in the literature. [2] Recent studies showed that nanostructuration of plutonium oxide appears to play a key role in the sonochemical formation of Pu colloids in aqueous solutions. [3] In this work, the sonochemical behavior of nanostructured thorium oxides is studied. First, different nanostructured ThO2 were prepared by oxalic route or in basic conditions in the presence of a polymer before being characterized. The sonochemical reactivity of these oxides was then studied in dilute aqueous solutions at low frequency ultrasound (20 kHz) under Ar/20%O2 bubbling. The effects of cavitation on solid materials were studied with Raman spectroscopy, AFM, SEM and HR-TEM, while the liquid medium was characterized by UV-vis and ICP-AES spectroscopies. Despite its very refractive behavior, a significant amount of ThO2 was found to dissolve in sonicated 0.5 M H2SO4 solution. In more dilute acidic conditions, a lower dissolution of ThO2 was observed. In these media, the sonolysis of nanostructured ThO2 allowed its conversion into a new crystalline phase. The formation of this phase can be attributed to the physical and chemical effects driven by the acoustic cavitation which allows: (i) the oxide particle fragmentation and size reduction, and (ii) the sonochemical generation of H2O2. This compound was then prepared in homogeneous solution in the absence of ultrasound. UV-Vis, Raman, FTIR spectroscopies coupled to synchrotron X-Ray Diffraction and EXAFS spectroscopy evidenced the formation of a thorium peroxo sulfate (Fig. 1), poorly referenced in the literature and for which a structure has never been reported.

Fig. 1 :

 1 Fig. 1: Fourier Transform of experimental EXAFS spectrum obtained for the Th peroxo sulfate.


Neutronic and fuel cycle comparison of uranium and thorium as matrix for minor actinides bearingblankets

Introduction:

Minor actinides are a set of three main elements (neptunium, americium and curium) which are byproducts of uranium irradiation in nuclear reactors. These elements are produced in relatively limited quantities [START_REF] Nea | Status and assessement of fission products and minor actinides partitionning and transmutation[END_REF] but they exhibit long-term radiotoxicity and decay heat levels which complicate the handling of associated nuclear waste.

In the case of a closed nuclear fuel cycle strategies where spent fuel is reprocessed and plutonium reused in fast spectrum reactors, minor actinides are the main contributors to long term radiotoxicity of the spent fuel and to decay heat of the ultimate waste package. Minor actinides transmutation has thus been proposed as a potential solution to decrease the burden of nuclear waste and to reduce the constraint on the final repository. [START_REF] Salvatores | Nuclear Waste Transmutation[END_REF] Transmutation in critical reactors can be done in two different ways:

-Homogeneous transmutation, in which minor actinides are directly mixed with the reactor fuel. This solution exhibits the best performances as the minor actinides are exposed to a high flux level. However, it exhibits the drawback of contaminating the entire fuel cycle with minor actinides and it decreases the "safety" performances of the reactor. Minor actinides content of up to 5 %vol can be loaded depending on the considered core design. Additionally, the residence time of the minor actinides bearing fuels can not exceed the one of standard MOX fuel. -Heterogeneous transmutation in which minor actinides can be loaded in specifically designed assemblies, usually in the periphery of the core, which are called "Minor Actinides Bearing Blankets" or MABB. The use of such subassemblies helps decoupling minor actinides management from the fuel and thus gives a larger flexibility compared to the homogeneous mode. As these blankets benefit from the neutron leakage from the active zone they have almost no impact on the core neutronic parameters such as delayed neutron fractions or sodium void worth. This allows to load a large minor actinide mass and to reduce the number of MABB to be manufactured. On the other hand, the obtained performances are lower than the previous one as the flux level seen by the assemblies is quite low. Minor actinides content between 10 % and 40 % are expected to be loaded in such cores. As fuel and MABB cycles are decoupled, higher transmutation rates can be expected at the cost of longer irradiation time.

The present paper focuses on heterogeneous transmutation strategies. A thorough analysis of this transmutation approach has been carried out by a NEA task force in 2009 and summarized in [START_REF] Nea | Homogeneous versus heterogeneous recycling of transuranics in fast nuclear reactors[END_REF]. The main points are described below:

-The high content of minor actinides in the fuel requires important fuel design effort, notably in terms of mechanical design. Previous experiments, especially the SUPERFACT experiment in which pins with up to 45 % of americium and neptunium were irradiated in the Phenix reactor core [START_REF] Prunier | Some specific aspects of homogeneous Am and Np based fuels transmutation through the outcomes of the superfact experiment in PHENIX fast reactor[END_REF] showed that MABB irradiation was accompanied by an important production of Helium due to alpha decay of minor actinides nuclei which has an impact on the mechanical behavior of the pin and on the size of the gas plenum. -Power production in the MABB assemblies is also very low at the beginning of irradiation which puts tighter constraints on the mechanical design as fuel restructuration does not happen at low temperatures. The important power variation during irradiation also increases the strain on the fuel pins, possibly leading to thermal cracking. [START_REF] Bonnerot | Development programme on minor-actinide-bearing blankets at CEA[END_REF] -Decay heat, gamma and neutron emission of irradiated and refabricated MABB assemblies is significantly higher than for a standard MOX fuel, which leads to additional issues in terms of fuel handling, reprocessing and manufacturing.

When considering minor actinides transmutation, several objectives are usually pursued. Firstly, the transmutation performances, e.g. the amount of minor actinides which are effectively turned into fission products during irradiation is considered, as it is a direct estimator of the performances of the process. The support ratio, which is the number of reactors which production can be absorbed in one minor actinide burner, is also of interest from an economic point of view. The support ratio can be reduced either by decreasing the production of minor actinides in the reactor or by increasing transmutation performances, as discussed here.

It has been proposed to add moderating material in the MABB in order to increase the transmutation performances of the design [START_REF] Saint | Americium once-through of moderated targets in a CAPRA core[END_REF]. Slowing down the neutrons in the blankets has the interest of increasing the absorption cross sections and thus the number of captures or fissions. However, it also increases the amount of curium produced, which is more troublesome than americium on fuel back-end due to a higher neutron source and specific decay heat. This addition is especially interesting in case of once-through transmutation, in which the blankets are irradiated only once and then discarded as waste. We considered here a heterogeneous transmutation scheme in which irradiated blankets are reprocessed to maximize amount of transmuted material.

Similarly to plutonium, minor actinides cannot be loaded per se as oxide fuel in a reactor but must be blended with a matrix to produce usable reactor fuel. Several materials have been proposed as potential matrices for MABB fuels. The first one is evidently uranium dioxide (UO 2 ), which has been tested in the SUPERFACT experiments for instance. UO 2 is a well-known material with a low swelling rate and which can withstand the long residence time associated with heterogeneous transmutation. However, the use of uranium oxide as support matrix comes with a production of plutonium which may cause an issue in terms of proliferation. It also implies a modification of the core to keep a total breeding gain close to unity. An analysis of the impact (or lack thereof) of the use of minor actinides blankets has been done in [START_REF] Buiron | Heterogeneous minor actinides transmutation on a UO2 blanket and on (U,Pu)O2 fuel in sodium-cooled fast reactor. Assessment of core performances[END_REF]. Additionally, the irradiation behavior of mixed uranium-minor actinides oxide fuel has yet to be fully characterized [START_REF]Status of minor actinides fuel development[END_REF].

Inert Fuel Matrix, or IMF, has also been discussed for transmutation in heterogeneous mode. A review can be found in [START_REF] Matzke | Materials research on inert matrices: a screening study[END_REF]. In this concept, the minor actinides are embedded as oxide in either a ceramic material (Cercer concept) or a metallic material (Cermet). This removes the production of plutonium due to capture by uranium 238 in the target but the selection of the matrix is complicated as it should exhibit a good thermal conductivity, acceptable swelling under irradiation and good irradiation resistance behavior to neutrons, alpha and fission products. No matrix has been found featuring all these parameters. However, a possible hybrid matrix of AmZrO 2 dispersed in an MgAl 2 O 4 matrix, which limits damage to MgAl 2 O 4 by fission products irradiation while making good use of its otherwise good stability has been proposed in [START_REF] Chauvin | Optimisation of inert matrix fuel concepts for americium transmutation[END_REF] . Issues regarding dissolution of the inert matrix must also be addressed (see for instance [START_REF] Ebert | Dissolution studies on molybdenum-based inert matrix fuels for the transmutation of minor actinides[END_REF]).

Thorium has been proposed as a potential nuclear fuel in the Th232/U233 fuel cycle, in which fissile uranium 233 is bred from thorium 232. This cycle can be closed in fast or thermal reactors, although it requires an initial stock of fissile material (U235 or Pu239) to start the breeding process. The potential benefits coming from the use of this cycle are listed in [START_REF]Thorium fuel cycle -Potential benefits and challenges[END_REF]. To name but a few, this option virtually removes minor actinides production and increases the reserve in fertile material by a factor three to four as thorium is more abundant than uranium while being intrinsically proliferation resistant due to high gamma production of daughter isotopes of U232. India is currently the leading country for thorium fuel cycle industrialization. Thorium dioxide (ThO 2 ) or thoria is also a relatively well-known material which performances under irradiation are better than those of UO 2 . A detailed study of thorium properties as a nuclear fuel can be found in [START_REF] Hania | Advanced Fuels/Fuel Cladding/Nuclear Fuel Performance Modeling and Simulation[END_REF] Thorium dioxide use has been already discussed as support matrix for heterogeneous transmutation in once-through scheme, for instance in [START_REF] Lombardi | Thoria and inert matrix fuels for a sustainable nuclear power[END_REF]. In this case, advantage is taken of the low solubility of thoria in groundwater for long-term storage. Additionally, this option limits the production of plutonium in the blankets, thus decreasing the total radiotoxicity at disposal. Advantage has also been taken of the lack of plutonium to transmute plutonium and minor actinides without separation during reprocessing, in the case of it being not acceptable. The use of thorium axial blankets in such a case and the related neutronic impacts are discussed in [START_REF] You | A neutronic study on advanced sodium cooled fast reactor cores withthorium blankets for effective burning of transuranic nuclides[END_REF].

We elaborate in this paper on the possibility of using thorium dioxide (ThO 2 ) as a support matrix for minor actinides bearing blankets in the case of a multi-reprocessing scheme in plutonium-fueled fast reactors. We compare the relative performances of uranium and thorium for this application in terms of reactor and fuel cycle impacts. The methodology and tools used are detailed in the first part and the effects of thorium and uranium matrixes on reactors parameters, fuel cycle and transmutation performances are then analyzed in the following sections.

Several cases can be envisioned for transmutation with a thorium support. We considered that thorium was used in combination with a conventional U/Pu fuel cycle. The following possibilities for thorium use which were investigated here are :

-Thorium could be used only as support matrix and the bred uranium 233 can be recovered after reprocessing and used for starting an independent thorium/uranium cycle. As this cycle requires an initial supply of uranium 233 for starting, this solution would allow a reduction of the total inventory of minor actinides during the switch to thorium while producing the necessary uranium 233. Similarly, the uranium production could be incorporated in the reactor core as fuel, thus replacing part of the plutonium and decreasing the minor actinides production. This option was not pursued here. -Uranium 233 produced during irradiation could be reused as a neutron supplier directly in the blankets, in order not to mix plutonium and uranium 233 in the standard fuel cycle. In this approach, plutonium from the blankets is recovered to be used in the reactor core while uranium, thorium and minor actinides are left in the blankets. The transmutation performances would then benefit from the presence of fissile material in the blanket without requiring plutonium for this purpose, thus limiting the increase in the plutonium inventory. -Finally, thorium could be used in conjunction with a reprocessing scheme which prevent recovery of potential proliferating materials [16] [17], in which case all the elements produced during irradiation would be recovered together and incorporated again into MABB. This case totally separates the management of the fuel cycle and of the blankets.

We compared those strategies with two similar ones using uranium dioxide as support matrix:

-The direct one in which UO 2 is used as matrix and plutonium is recovered during reprocessing for use in the reactor core. Minor actinides are then reincorporated in fresh blankets. This strategy is currently envisioned in France as discussed in [START_REF] Meyer | Scenarios for Minor Actinides Transmutation in the Frame of the French act for waste management[END_REF]. -The TRU approach in which there is a complete separation of the driver fuel cycle from the blankets fuel cycle. This approach is promoted by the USA in which plutonium and minor actinides are not separated in order to remove any proliferation issues. Plutonium is considered as a minor actinide in this case and treated as such. An example of implementation of this strategy is discussed in [START_REF] Bays | Heterogeneous transmutation sodium fast reactor[END_REF].

Methodology

Neutronic calculations were performed using the ERANOS code package [START_REF] Rimpault | The ERANOS code and data system for fast reactor neutronic analyses[END_REF]. The starting point of the calculations was a critical homogeneous SFR core developed jointly by CEA, EDF and Areva and detailed below in Table 1. The core was described using 2D-RZ geometry with a homogeneous description of the various core mediums. Calculations were based on the nuclear data library JEFF 3.1 [START_REF] Santamarina | The JEFF-3.1.1 Nuclear Data Library[END_REF]. A layout of the core is given in Figure 1. The plutonium content was tuned to fulfill a reactivity criteria at the end of cycle (ρ FC = 750 pcm). Calculations were done using with the transport equation with a 33-group energy mesh. The design of the target assembly and fuel pin is still underway so the analysis was done with 38 % in volume of heavy metals in the target assemblies, which is deemed representative of the expected result. Transmutation performances were evaluated using the two following estimators:

-Transmutation rate calculated between the beginning of irradiation and after 5 years of cooling and 2 years of manufacturing :

--Support ratio, which is defined as the ratio of the minor actinides consumption in the blankets with 20 % MA loaded, divided by the production in the reactor core alone. This indicates the number of reactors which minor actinides production can be transmuted using one reactor equipped with radial blankets. Minor actinides loading was done with a minor actinides vector hereafter labelled "MA2035" and given in Table 2. This isotopic vector is deemed representative of the minor actinides feed available by 2035 in France considering the foreseen evolution of the French PWR fleet. The plutonium vector used for the calculations, given in Table 3 was also considered representative at the same time scale. Minor actinides bearing assemblies were loaded in the first outer ring of the core in substitution of the first reflector ring. Their residence time was twice the one of standard fuel assemblies in order to compensate for the associated low flux level. This amounted to 4100 EPFD, i.e. 10 reactor cycles. In the uranium matrix case, at the end of the irradiation, blankets were reprocessed with 5 years cooling time and 2 years manufacturing time. The plutonium was removed from the blankets and mixed with the core production. Uranium is reused as matrix for the blankets. The minor actinides content in the blankets was recompleted using the minor actinides vector coming from the core production in order to keep the loaded mass constant over two cycles. The external feed of minor actinides is assumed to come from similar reactors without radial blankets. Mass balance is then achieved using uranium. This strategy will be called as U/Pu/MA in the following. It is shown in Figure 2. In the thorium case, blankets were reprocessed on the same time scale with the plutonium being recovered and added to the plutonium of the core. The uranium produced during irradiation, mainly uranium 233, is taken out of the blankets while minor actinides are recompleted using the minor actinides vector from the core. This strategy will be called Th/U-Pu/MA in the following. The uranium here produced could be mixed with plutonium and used as fuel but this option was not investigated here. It is also shown in Figure 2.

A third strategy was investigated in which thorium is used as support matrix and uranium is left in the blankets along with minor actinides and plutonium is removed for use in the reactor core. This strategy is denominated Th(U)/Pu/MA. The feasibility of thorium/U/Pu separation still yet has to be demonstrated, as discussed in [START_REF]Thorium fuel cycle -Potential benefits and challenges[END_REF], so it is possible that this strategy may incur penalties in terms of reprocessing losses.

A final situation in which the blankets are reprocessed on a Th(U)/TRU basis was also studied. In this case, uranium, plutonium and minor actinides from the blankets are kept in the blankets. Minor actinides with the core isotopic composition are then added to the blankets up to 20 % of minor actinides and mass balance is achieved with thorium. A similar strategy named U/TRU was also studied, which is similar to the Th(U)/TRU one, with all the transuranics being kept in the blanket. As said before, this strategy completely separates the driver and the blankets fuel cycles and reduces the proliferation risks. All these strategies are shown in Figure 3. The whole system was taken to equilibrium and characterized both at the end of the first 10 irradiation cycles and at equilibrium. The equilibrium is reached when the transmutation performances over two consecutive cycles are identical. We denominated as EOEC the end of a MABB equilibrium cycle.

The minor actinides content is stalled to 20% which corresponds to 2375 kg of minor actinides (heavy nuclides mass) in both cases. For perspective, the current minor actinides production of the French nuclear fleet amounts to 1.2 tons a year. The minor actinides isotopic vector of the core production is used as make-up feed to reach the 20 % limit which is currently considered by CEA [START_REF] Buiron | Heterogeneous minor actinides transmutation on a UO2 blanket and on (U,Pu)O2 fuel in sodium-cooled fast reactor. Assessment of core performances[END_REF].

The use of thorium as matrix material was analyzed through its impacts on the core parameters such as breeding gain, transmutation performances, sodium void worth and Doppler Effect and on the fuel cycle with neutron and gamma source, decay heat, radiotoxicity estimated.

Comparison of uranium and thorium as support matrix for MABB : transmutation performances a. Transmutation rate

The first parameter of interest with regards to transmutation performances is the transmutation rate, which can be related to the efficiency of the transmutation process. Several comments can be made on Figure 4. The options U/Pu/Ma and Th/U-Pu/MA, which correspond to the case where only the minor actinides are left in the blankets, yield the lowest transmutation rate. Indeed, for these two strategies, no fissile material is initially present in the blankets. Consequently, the neutron production in situ is limited compared to the other cases, which in turn limits the transmutation performances. The thorium case is slightly lower than the uranium one as the thorium mass in the blankets is inferior to the uranium mass due to the thoria density being only 10 g.cm -3 compared to 10.95 g.cm -3 for uranium dioxide. This leads to a lower production of fissile material during irradiation and thus a slightly lower transmutation rate.

In all the three other cases, fissile materials are left in the blanket during reprocessing. The transmutation rate is then directly proportional to the amount of fissile material in the blanket. For the Th(U)/Pu/MA strategy, the fraction of U233 in the blanket is 7 % and for the two TRU cases, volume fractions of "fissile" material (U233, U235, Pu239 and Pu241) at the beginning of irradiation at equilibrium are respectively 7.8 % and 8.0 % for Th/U(TRU)) and Th(TRU). The increase in the transmutation performances is directly linked to the increase in fissile material in the blanket assembly. The latter has two mutually opposing effects on the neutron spectrum in the blankets. On the one hand, the increase in the flux level leads to a direct increase in the reaction rate, while on the other hand, the spectrum hardening leads to a decrease in the absorption cross section, thus decreasing the transmutation rate.

To evaluate the magnitude of each effect, the same calculations were carried out over 410 EPFD using a simplified depletion equation while varying in turns the spectrum and the flux level. Two calculations were performed with the neutron spectrum from the U/Pu/MA case and two different flux levels and the same jobs were performed again with the Th(U)/Pu/MA spectrum. Looking at Table 5, we can see that the impact of the spectrum hardening is limited to around -0.1 % point in the transmutation rate. Looking at the impact of the flux level, we can see it is dominant compared to the spectrum effect, and that is this effect which explains the increase in transmutation performances.

b. Support ratio

The results for support ratio evaluation are given in Table 6. Similarly to what can be seen on Figure 4, obviously the strategies where fissile materials are left in the blankets are more effective than the two strategies without fissile material. We can also see that the U/TRU approach yields the best support ratio. As the core production of minor actinides is nearly not impacted by the presence of minor actinides bearing blankets, the support ratio is proportional to the transmutation rate. The impacts of each strategy on core operation parameters are given in Table 7. The reactivity swing in the thorium case is higher at the beginning of irradiation than for uranium as thorium is more capturing than uranium and the build-up of plutonium 239 and uranium 233 occurs in the first cycles. At equilibrium, the reactivity swing is higher for the case with fissile material in the blankets as these blankets act like fuel assemblies. Conversely, the average burn-up is lower as power production is spread over one more ring of assemblies.

The cases where thorium and uranium 233 are kept in the blankets have a lower impact on the core sodium void worth as the -factor (neutron production per fission vs energy) of U233 dependency on the neutron energy is flatter than to the one of plutonium at high energy. Consequently, in case of coolant voiding, the contribution of the blankets to the total sodium void worth is lower. The case with uranium leads to a higher power production in the blankets by 8 % at equilibrium which is due to the higher content in fissile material in this case. During the first cycles, both cases are equivalent. In both fissile-free cases (U/Pu/MA and Th/UPu/MA), the breeding gain is increased by 0.1 due to the presence of fertile blankets. In the other case, the breeding gain only increase by a factor six to eight as fissile material is already present in the blankets. This residual increase is mainly explained by production of more fissile isotope such as plutonium 238 during the transmutation process. The evolution of the heavy nuclides content in the blankets is given in Figure 6. The behavior of each element can be divided into two consecutive patterns constituted of a first irradiation phase of 4100 EPFD followed by a second one of similar-length. Then, the blankets composition is evolved for 2555 days to simulate reprocessing and the blankets are recompleted using the minor actinides vector from the core production (also after 2555 days of decay) to achieve 2375 kg of MA in the blankets.

Mass balance is achieved using the support element. Cooling time was omitted on Figure 6 for clarity.

Thorium and uranium in the cases where they are used as support matrix behave similarly. Their mass decreases during irradiation and are stable during decay phase. Both the time necessary to reach an equilibrium situation and the amount of thorium/uranium in the blankets depends on the amount of fissile elements in the blankets. Both Th/U-Pu/MA and U/Pu/MA reaches equilibrium after on complete MABB cycle (10 reactor cycles). Cases where fissile material is present in the blankets require a longer time to achieve equilibrium. Regarding uranium evolution, it should be noted that the uranium mass in the Th(U)/TRU or Th(U)/Pu/MA is remarkably stable during irradiation, which will have a positive effect on power variation in the blankets during irradiation. Considering plutonium production, it is interesting to observe that half of the plutonium production in the U/Pu/MA case is coming from minor actinides transmutation, especially from the reaction . Using thorium as support matrix consequently halves the production of plutonium in the blankets and the related inventory. Regarding the evolution of minor actinides mass in the blankets, the behavior of each curve on Figure 6 is directly related to the transmutation rate plotted on Figure 4.

The uranium 233 vector is slightly different between the Th(U)/Pu/MA and Th(U)/TRU cases with respectively 63 and 53 % of uranium 233. The exact composition is given in Figure 7 at EOEC. It can be seen that the fraction of U232, which has a strong gamma emitter in its decay chain is the same in both cases. The plutonium isotopic vector is however widely varying between the various cases. As it can be seen in Figure 8, the Pu239 content in the thorium cases is very limited, making the plutonium useless for reactor use without dilution with higher-grade plutonium. This plutonium could for instance be used as denaturing material for military grade plutonium [START_REF] Kessler | Plutonium denaturing by Pu-238[END_REF] or mixed with high-grade plutonium depending on the considered limits for proliferation resistance. It should be noted that the total mass of plutonium 238, which is the main responsible for decay heat at the reprocessing time-scale is similar between the two cases, as this production is related to the transmutation process and not to the breeding on the support. Looking at Table 8, the final inventory in minor actinides is directly linked to the transmutation performances of the strategy considered. The strategies with the highest transmutation rates also lead to a shift of the isotopic vector towards heavier curium isotopes, both in terms of mass and in terms of content. This shift is due to the higher level of flux which permits successive capture on curium isotopes and to the reprocessing of curium in the minor actinides. This can be seen while comparing Th/U-Pu/MA and U/TRU strategies for instance. The effect of this shift on fuel cycle parameters such as decay heat or neutron source is evaluated subsequently in this work.

c. Power levels in the blankets

Power level variation in the blankets is due to the build-up of fissile isotopes during irradiation. One part is due to production of fissile material such as Am242m from the minor actinides initially loaded and the other part is explained by the support. Figure 9 shows the evolution of the power level in the blankets. For fissile-free strategies, we can observe an increase in the power level in the blankets due to the production of plutonium 239 or uranium 233. The sharp increase observed after five cycles is explained by the calculation scheme used, which adjusts the enrichment of the entire core to take into account the fissile material in the blankets, thus increasing the flux region. This also explains the behavior of the three other strategies.

Looking at Figure 9, it can be observed that the cases Th/U-Pu/MA and U/Pu/MA are very similar in terms in power level and power variation during irradiation. This is explained by the production of Pu239/U233 during the first phase of irradiation followed by a quasi-equilibrium state during the second irradiation phase. The minor actinides contribution is similar for each case. Power variation in the region is 144 MW over ten reactor cycles for the thorium case and 113 MW for the uranium one, which corresponds respectively to a 218 % and 152 % increase in blanket power. For the TRU cases, the power level can reach up to 720 MW due to high fissile content in the blankets. This is equivalent to adding a new ring of fuel assemblies in the periphery of the core. In the case, the blankets assemblies are equivalent to fuel assemblies in terms of power at the end of irradiation but undergo a 45 % increase in their power from 400 to 720 MW, compared to a 20 % maximal increase for standard fuel assemblies. This high power variation may have adverse effects on fuel mechanical behavior, especially for swelling and fission gases release behavior [START_REF] Bonnerot | Development programme on minor-actinide-bearing blankets at CEA[END_REF].

The Th(U)/Pu/MA strategy is an intermediate one with a higher power level which may be beneficial for pins mechanical behavior compared to fissile-free strategy. The lower power variation limited to 193 MW or a 156 % increase, may also limits the constraints on the assembly design. As such, it may be interesting as a solution which could increase the transmutation performances without hindering assembly conception.

Comparison of uranium and thorium as support matrix for MABB : impact on fuel cycle parameters

In terms of fuel cycle, the estimators that were selected are the decay heat, gamma source and neutron source of the spent fuel. R&D is currently still required to ascertain the feasibility of handling, reprocessing and manufacturing of minor actinides bearing blankets and we showed here that using thorium instead of uranium does not add new requirements to the fuel cycle part. The depletion calculations were performed using the DARWIN code package [START_REF] Tsilanizara | DARWIN: An Evolution Code System for a Large Range of[END_REF].

a. Manufacturing and reprocessing of thorium oxide fuels An extensive review of the benefits and challenges associated with the thorium fuel cycle has been done by the IAEA in [START_REF]Thorium fuel cycle -Potential benefits and challenges[END_REF]. Generally speaking, the main challenge associated with thorium reprocessed is the industrial implementation and validation of the THOREX process developed at Oak Ridge in the fifties. This implementation requires the development of adequate shielding solutions to overcome the issue of high gamma emissions from uranium 232 daughter nuclei such as thallium 208, thorium 228 or lead 212. The presence of hard-to-extract protactinium 233 in the spent fuel is not an issue since its 27 days period means it has completely disappeared from the spent fuel after five years of cooling. A second challenge is the development and industrial validation of a process which can successfully separate and recover U, Pu and the minor actinides from the spent fuel.

It should be noted here that the plutonium obtained from the separation of plutonium from thorium during recycling of the blankets may not be suited for military applications as its Pu 238 fraction is too high according to [START_REF] Kessler | Plutonium denaturing by Pu-238[END_REF] , as shown in Figure 8.

b. Back end : Decay heat and neutron source Decay heat per assembly is a dimensioning factor for the reprocessing flowsheet. A "hotter" fuel assembly will require extended cooling time before handling is possible or more expensive handling devices. Similarly, an equilibrium between longer cooling time and decreased reprocessing performances will have to be found as more active fuels degrades reprocessing efficiency [START_REF] Gao | Influence of alpha and gamma radiolysis on Pu retention in the solvent TBP/kerosene[END_REF]. Time evolution of decay heat in the spent blanket assemblies can be separated in two steps as seen in Figure 10. The first one, from end of irradiation to about two years is characterized by a rapidly decreasing decay heat mainly fueled by short-lived fission products and curium 242 decay. The U/TRU strategy is the less penalizing in terms of decay heat for this timescale as in this case curium 242 concentration has already peaked and actually started decreasing under irradiation. On the longer time scale, from two years to a century, the decay heat decrease is slower as the main contributors are Pu238 and Cm244, with half lives of 87.8 and 18.1 years. The decay heat for the Th(U)/Pu/MA strategy is the lowest for this time scale, as more americium 241 has been consumed than in the fissile-free cases and less curium has been produced than in the TRU cases. Irradiated assemblies in the Th(U)/Pu/MA are on the average 10 % less active than those of Th(U)/TRU, which is the most penalizing case.

If the impact of the strategy on decay heat is relatively limited, neutron source can change up to 30 % between the two extremal cases. Neutron source is mainly linked to curium isotopes heavier than curium 244 and is key factor for evaluating transportation and reprocessing feasibility, with regards both to radioprotection and criticality safety. In the U/TRU case, the plutonium accumulation in the blankets creates a reservoir for americium production and curium production onwards, which explains the higher neutron source associated with this strategy. In this case, using thorium as a support matrix leads to a reduction of 10 % in the total neutron source on the short term. In both fissile-free strategies, the neutron source remains lower than 6e10 n/s/assembly due to a low production of higher curium isotopes as it can be seen in in Table 8. c. Gamma emission A potential issue associated with the use of thorium is the increased gamma emission coming from thorium 228 and its daughter-nuclei or Pa 233, which are strong gamma emitters [START_REF]Thorium fuel cycle -Potential benefits and challenges[END_REF]. The level and spectrum of gamma emission 30 days after the end of irradiation (assembly extraction of the core for relocation) and 5 years (typically expected cooling time) have been evaluated, to assess the shielding requirements both for handling the irradiated assembly during refueling operations and during reprocessing. The results are given in Figure 12 and Figure 13. A comparison of the total level of gamma emissions in each case is also given in Table 9. It should however be noted that due to the high content in minor actinides, remote handling will be necessary both for manufacturing and reprocessing of transmutation targets. Concerning Figure 12, two remarks can be done. First, the presence of low energetic gammas from Protactinium 233 can be observed along with a small contribution from Thallium 208 in the 2 MeV range. These two contributions are directly due to nuclei produced by capture on thorium or thorium disintegration chain and are thus not visible in the spectrum for the U cases. The second point to make is that several peaks such as for Cs 137 or Ce 141 are corresponding to fissions products. The height of these peaks is linked to the amount of fissile material in the blankets and to the transmutation performances of the strategy.

Contrary to the decay heat and neutron source, it is expected that the evolution kinetic of gamma emission will be different between the two cases as the main emitters are not the same. However, as most of the gamma emissions (more than 70 %) come from fissions products which are roughly the same regardless of the support matrix, the difference in evolution time-scale is limited, as it can be seen on Figure 14. Regardless of the strategy, the total gamma emission of the blankets assembly is always lower than the one of a standard fuel assembly, as the one of such an assembly is dominated by fission products.

d. Conclusion

The conclusion of this analysis is that the use of thorium instead of uranium as support matrix for MABB will have a limited impact on spent fuel neutron source and decay heat, which means that no additional constraints have to be considered on the conditioning of the waste for final disposal. However, the increased number of fissions in the blanket assemblies in the thorium and TRU cases means that more fission products are going to be found in the assemblies to be reprocessed and thus the gamma emission of these elements will be higher compared to the fissile-free cases. Thorium itself was not found to have a significant impact on the gamma emission of the fuel elements except between 30 and 300 days where Protactinium 233 is responsible for up to 30 % of the emission. The exact impact of the 2.6 MeV gamma emitted by Thallium 208 on the fabrication and retreatment processed should be more thoroughly assessed in the future, notably in terms of additional shielding required. However, the high minor actinides content is likely to induce the need for adequate shielding, thus mitigating the increase in safety requirements for handling thallium 208.

Conclusion

We can compare the various strategies with the U/Pu/MA, which is the one currently envisioned in France [START_REF] Varaine | Heterogeneous recycling in SFR core periphery[END_REF]. This is done in Table 10. We can see that simply replacing uranium with thorium in the blankets and extracting the uranium 233 for use elsewhere yields performances lower than in the reference case, except for the impact on reactor operation due to the flatter -value of the uranium 233 compared to Pu239.

Uranium 233 can also be left in the blankets to fasten the transmutation process. In this case, we compared a strategy where only uranium 233 is left to two strategies where all the transuranics elements are left in the blankets with either uranium or thorium support. Again, the thorium impact on core operation is lower. In terms of fissile inventory, the U/TRU case leads to the highest fissile inventory in the blanket while use of thorium leads to a 10 % lower inventory. The neutron sources at 30 days and at five years are also lower in the thorium case due to a lower of production of heavy curium isotopes. In terms of decay heat, thorium is less effective at 30 days but leads to a lower long-term decay heat source. It can be concluded that the use of thorium as a substitute for uranium support in minor actinides bearing blankets does not yield increased transmutation performances if the uranium 233 thus produced is extracted for use outside the blankets. On the other hand, the use of thorium blankets with a 'TRU' scheme in which all the transuranics, plutonium included, are left in the blankets for further irradiation yields similar performances to the uranium one albeit with a lower impact on the core parameters due to better neutronic behavior of uranium 233 and a lower overall fissile inventory by 16 %. The intermediate option of reusing only the uranium 233 produced in the blankets to speed up the transmutation process yields intermediate performances but require fine reprocessing of the irradiated fuel. This approach also increases the transmutation performances compared to a fissile-free approach while limiting the power variation in the blankets which is favorable in terms of assembly design. We can conclude from this work that the choice of thorium instead of as support matrix for heterogeneous transmutation targets cannot be decided in terms of transmutation performances only but requires careful analysis of the potential gains it can bring with regards to the overall fuel cycle.

Figure 1 :

 1 Figure 1 : 2D-RZ representation of the SFR-V2B core with minor actinides bearing blankets

Figure 2 :Figure 3 :

 23 Figure 2 : Outline of the Th(U)/Pu/MA, U/Pu/MA and Th/U-Pu/MA strategies

Figure 4 :

 4 Figure 4 : Evolution of the transmutation rates in the MABB

  -01 4.64E-02 Pu239 5.77E-01 1.86E+00 Pu239 5.32E-01 1.86E+00 Np237 1.72E+00 3.07E-01 Np237 1.55E+00 3.50E-01 Am241 1.96E+00 2.51E-01 Am241 1.81E+00 2.89E-01 Am243 1.76E+00 1.83E-01 Am243 1.60E+00 2.12E-01 Cm244 9.14E-01 3.88E-01 Cm244 8.31E-01 4.38E-01

Figure 5 :

 5 Figure 5 : Difference between the spectrum in the blankets in the thorium case and in the uranium case

Figure 6 :

 6 Figure 6 : Evolution of the Th, U, Pu and MA content in the blankets vs reactor cycle. Cooling and manufacturing were not plotted.

Figure 7 :

 7 Figure 7 : Uranium vector at equilibrium loading for Th(U)/Pu/AM and Th(U)/TRU strategies

Figure 8 :

 8 Figure 8 : Plutonium isotopic vector at equilibrium for U/Pu/MA, Th(U)/PuMA, U/TRU and Th(U)/TRU. The case Th/UPu/MA is similar to the Th(U)/Pu/MA in terms of isotopic vector.

Figure 9 :

 9 Figure 9 : Evolution of the power level in the blankets during irradiation

Figure 10 :

 10 Figure 10 : Evolution of decay heat per assembly for each strategy

Figure 11 :

 11 Figure 11 : Evolution of neutron source vs cooling time for each strategy

Figure 12 :Figure 13 :

 1213 Figure 12 : Gamma spectrum after 30 days of cooling of irradiated CCAM assembly

Figure 14 :

 14 Figure 14 : Comparison of the evolution of gamma emissions for three cases

Table 1 : Characteristics of the SFR V2b core used for the calculations [22]

 1

	Main characteristics of SFR V2b
	Power (MWTh/MWe)	3600/1450
	Number of fissile assemblies (Internal core/External core)	453(267/186)
	Number of assemblies in the first outer ring (MABB)	84
	Residence time	5 x 410 = 2050 EPFD
	Volume fractions	43.7 % UPuO 2, 27.5 % Na, 8.2 %
		Wrapper, 11.9 % cladding
	Fuel nature	UPuO 2
	Plutonium content(mean/IC/EC) %vol	15.80%/14.65%/17.44%
	Effective breeding ratio	1

Table 2 : Isotopic vector used for minor actinides

 2

	Element	Np237 Am241 Am242m Am243 Cm242 Cm243 Cm244 Cm245 Cm246
	Fraction	16.87 60.62	0.24	15.7	0.02	0.07	5.14	1.26	0.08
	(%mass)
	Isotope	Pu238	Pu239	Pu240		Pu241	Pu242		Am241
	Fraction	3.57	47.39	29.66		8.23	10.37		0.78
	(%mass)

Table 3 : Isotopic vector used for plutonium

 3

Table 4 : Comparison of the one-group cross sections between the uranium and thorium case

 4

Table 5 : Comparison of the effect of cross sections variations and flux levels on transmutation rate over one cycle

 5

		Flux level of Uranium case : 9e14	Flux level of Thorium case : 1,2e15
	Transmutation rate	n/cm²/s		n/cm²/s
	(%)	Uranium	Thorium	Uranium	Thorium
		Spectrum	Spectrum	Spectrum	Spectrum
	Support U	-5,2	-5,12	-6,39	-6,28
	Support Th	-5,27	-5,19	-6,43	-6,33

Table 6 : Support ratio for each strategy at equilibrium

 6

	Strategy	Th/U-Pu/MA U/Pu/MA Th(U)/Pu/MA Th(U)/TRU U/TRU
	Support ratio 1.61	1.77	2.2	2.55	2.68
	3. Comparison of uranium and thorium as support matrix for MABB : impact on core
	parameters
	a. Core operation parameters

Table 7 : Impact of the support on core parameters (MA 2035 vector, no moderating material, FC = fifth cycle, EQ = equilibrium cycle)

 7

		SFR
	Parameters	V2B Core (453	Th/U-PU/MA	U/Pu/MA	Th(U)/Pu/MA Th(U)/TRU	U/TRU
		S/A)
	Reactivity swing (FC) (pcm) Reactivity swing (EQ) (pcm)	-450	-518 -153	-419 -108	-518 -151	-518 -240	-419 -241
	Mean Burn-up, FC (GWd\tHM) Mean Burn-up, EQ (GWd\tHM)	99	97 93	97 93	97 87	97 83	97 82
	Max DPA (FC) Max DPA (EQ)	149	156 150	156 149	157 149	157 152	156 155
	Sodium void worth (pcm) (FC) Sodium void worth (pcm) (EQ)	1790	1808 1814	1818 1821	1808 1791	1808 1783	1820 1837
	Doppler Effect
	Magnitude (pcm)		-430	-440	-430	-430	-434
	(FC)	-441
	Doppler effect (pcm) (EQ)		-444	-448	-443	-428	-424
	Breeding gain
	(total/blankets)		0.088/0.093 0.086/0.082 0.088/0.093 0.088/0.093 0.086/0.082
	(FC)	0.009
	Breeding gain (total/blankets)(EQ)		0.106/0.039 0.103/0.058 0.078/0.029 0.063/0.008	0.062/-0.001
	Fraction of power
	in blankets at EOC	0	2.1	2.4	2.1	2.1	2.4
	(%) (FC)
	Fraction of power
	in blankets at EOC	0	5.4	5.64	10.5	13.5	14.7
	(%) (EQ)
	b. Inventory in the blankets

Table 8 : Minor actinides isotopic vector comparison at end of irradiation cycle at equilibrium

 8

	mass in kg	Th/U-Pu/MA	U/Pu/MA Th(U)/Pu/MA	Th(U)/TRU	U/TRU
	Np237	211.2	225.8		175.2		137.8	124.4
	Am241	694.9	667.0		543.4		458.1	401.1
	Am242m	69.5	70.1		57.4		49.6	43.8
	Am243	225.3	203.8		176.1		187.6	183.7
	Cm242	0.2	0.2		0.2		0.1	0.1
	Cm243	2.5	2.6		2.8		2.9	2.9
	Cm244	140.2	134.8		135.4		145.7	157.6
	Cm245	48.2	47.0		45.3		47.3	52.5
	Cm246	36.6	35.9		31.7		31.1	34.5
	Cm247	7.5	7.4		6.0		5.7	6.3
	Cm248	5.4	5.7		4.8		4.5	4.9
	Total	1441.5	1400.3	1178.2		1070.4	1011.7
	frac in %	Th/U-Pu/MA	U/Pu/MA	Th(U)/Pu/MA	Th(U)/TRU	U/TRU
	Np237	14.65		16.12			14.87	12.87	12.29
	Am241	48.21		47.63			46.12	42.80	39.64
	Am242m	4.82		5.01			4.87	4.63	4.33
	Am243	15.63		14.55			14.94	17.53	18.16
	Cm242	0.01		0.01			0.01	0.01	0.01
	Cm243	0.17		0.19			0.24	0.27	0.28
	Cm244	9.73		9.63			11.49	13.61	15.58
	Cm245	3.35		3.36			3.84	4.42	5.18
	Cm246	2.54		2.57			2.69	2.90	3.41
	Cm247	0.52		0.53			0.51	0.53	0.62
	Cm248	0.37		0.41			0.41	0.42	0.48
	Total	100.00		100.00		100.00	100.00	100.00

Table 9 : Comparison of the total level of gamma emissions at the end of irradiation

 9

	Strategy	Th/U-Pu/MA	U/Pu/MA Th(U)/Pu/MA	U/TRU	Th(U)/TRU
	30 days (W/assembly)	2559	1536	3680	3441	4081
	5 years (W/assembly)	89	95	208	327	289

Table 10 : Global comparison of the five strategies previously discussed

 10

	Strategy	U/Pu/MA (REF)	Th/U-Pu/MA	Th(U)/Pu/MA	Th(U)/TRU	U/TRU
	Transmutation rate (%)	39	37.4	48.5	53.3	56
	Sodium void
	worth (Δ with standard core	+31	+24 (-22.6 %)	+1 (-96.8 %)	-7 (-122 %)	+47 (+153 %)
	value, pcm)
	Reactivity
	swing (% of the standard	24	34 (+41 %)	34 (+41 %)	53 (+121 %)	54 (+125%)
	core)
	Power in
	blankets	5.64	5.4 (+4.4 %)	10.5 (+86.2 %) 13.5 (+139 %) 14.7 (+161 %)
	(MW)


M Virot

E Dalodiere

T Dumas

C Tamain

O Dieste Blanco

T Wiss

L Venault

P Moisy

Sergey I Nikitenko

Evidence for the Formation of a New Water Soluble Pu Peroxo Complex

établissements d'enseignement et de recherche

Evidence for the Formation of a New Water Soluble Pu Peroxo Complex

M. Virot, 1 E. Dalodière, 1 T. Dumas, 2 C. Tamain, [START_REF] Connick | [END_REF] O. Dieste Blanco, [START_REF] Fugate | Separations for the Nuclear Fuel Cycle in the 21st Century[END_REF] T. Wiss, 3 L. Venault, 2 P.

Moisy, 2 S. I. Nikitenko 1 The resolved peroxycarbonate complexe concerns a Pu(IV) dimeric unit coordinated by two side-on peroxide ligands. [4] In addition to the known peroxo complex of Pu(IV), the current work demonstrates that the addition of hydrogen peroxide to slightly acid Pu(IV) aqueous solutions may lead to the observation of a green solution exhibiting characteristic absorption bands at 455 and 660 nm which has, up to our knowledge, never been reported in the literature. The formation of this specie can also be observed from the sonochemical oxidation of 1 mM Pu(III) solutions under Ar/O 2 atmosphere (203 kHz, 250 mL, 20°C, 0.13 W mL -1 ). The green solution was characterized with Vis-NIR absorption spectroscopy, Raman spectroscopy, and XAFS techniques. We demonstrated the formation of a new polynuclear peroxo complex of Pu(IV) involving several Pu-O distances and at least 4 Pu atoms (Fig. 1). This water soluble complex was found to be stable for several months before it slowly converts into a Pu(IV) colloidal suspension exhibiting the characteristic absorption band at 620 nm and a strong Mie scattering in the near UV-vis range. HR-TEM and XAFS characterizations allowed to describe these colloids as monodispersed and quasi-spherical crystalline PuO 2 particles covered by hydrolyzed Pu(IV) moieties of approximately 2.4 nm.

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  oral  poster Hydrogen peroxide has been considered as an important reactant in nuclear industry for the stabilization of plutonium intermediate redox states or the precipitation of Pu peroxides used as precursors for PuO 2 preparation. The formation of two different Pu(IV) peroxo complexes, known as brown and red complexes, was reported to occur during the continuous addition of H 2 O 2 to Pu(IV) acid solutions.[1,2] Since the proposed structure given by Connick and W. H. McVey in 1949, only one Pu peroxo structure has been recently proposed in alkaline conditions.[1-3]

Fig. 1 :

 1 Fig. 1: Fourier transform of the k3-weighted EXAFS spectrum exhibiting two coordination spheres for the green peroxo complex.