Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

Journal of Nuclear Materials 462 (2015) 230–241

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Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH M. Magnin ⇑, C. Jégou, R. Caraballo, V. Broudic, M. Tribet, S. Peuget, Z. Talip Commissariat à L’Energie Atomique (CEA/DEN/DTCD), Service d’Etude et Comportement des Matériaux de Conditionnement, Marcoule BP 17171, Bagnols sur Cèze, France

h i g h l i g h t s  Oxidizing dissolution mechanism of MOX fuel.  Effect of the influence of the interim storage conditions.  Raman spectroscopy characterizations.  Precipitation of Studtite-type secondary phases.  Heterogeneous microstructure of the (U,Pu)O2 oxide.

a r t i c l e

i n f o

Article history: Received 6 November 2014 Accepted 11 March 2015 Available online 27 March 2015

a b s t r a c t The (U,Pu)O2 matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5–5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external c irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO2+ 2 ) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9  107 mol/l) were lower than for the non-irradiated reference experiment (1.4  105 mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO2 phase surrounding the Pu-enriched aggregates had not been particularly impacted by any major structural change compared to the data obtained prior to leaching. For the plutonium, its behavior in solution involved a continuous release up to concentrations of approximately 3  106 mol L1 with negligible colloid formation. This data appears to support a predominance of the +V oxidation state for plutonium in solution under highly oxidizing conditions. Furthermore, the Raman spectroscopy monitoring of the sample surface oxidation states did not point to any significant effect from the high Pu content of the aggregates (10–15%) and therefore did not indicate a better aggregate stability under radiolysis compared to the mainly UO2 matrix. This is because acidic pH conditions do not favor the development of oxidized layers on a fuel surface, with the exception of secondary phases. Ó 2015 Elsevier B.V. All rights reserved.

1. Introduction ⇑ Corresponding author at: Commissariat à l’Énergie Atomique (CEA), Marcoule Center, DTCD/SECM/LMPA, BP 17171, F-30207 Bagnols-sur-Cèze Cedex, France. Tel.: +33 466 397290; fax: +33 466 397708. E-mail address: [email protected] (M. Magnin). http://dx.doi.org/10.1016/j.jnucmat.2015.03.029 0022-3115/Ó 2015 Elsevier B.V. All rights reserved.

While awaiting reprocessing or direct geological disposal, MOX fuel assemblies may be stored underwater in special storage pools for several decades. Under such conditions, one incident scenario

M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241

to consider is that of a defect in the fuel cladding which could mean direct contact between the water and the fuel matrix. Even if the chemistry of storage pool water has proven to be relatively simple, the irradiation conditions expected (high c dose rate of the fuel assembly) and the presence of slightly acidic (pH 5–5.5) aerated water and dissolved oxygen would favor water decomposition and the formation of hydrogen peroxide (H2O2). This is likely to accentuate alteration of the fuel matrix at the defect point through oxidizing dissolution and the precipitation of possible secondary phases. Many studies have been carried out on Uranium Oxide-type irradiated fuels (UOX: UO2 95%) under oxidizing conditions, and have enabled descriptions of their alteration mechanisms under water. In the absence of any complexant in the solution, UO2 oxidation begins with the formation of an oxidized phase on its surface, the nature of which depends on the extent of the oxidizing conditions and on the solution chemistry [1–3]. Electrochemically, the process can be divided into two distinct stages: (i) a transitory stage during which the surface is oxidized from UO2 to UO2.33 (U3O7); (ii) above 145 mVSHE an oxidative dissolution occurs at a constant rate from a surface layer of UO2.33 (5–8 nm thick). Furthermore, alteration solution radiolysis leads to the formation of oxidizing species including H2O2 which is considered to play an important [4–6], and even major [7–9], role in the UO2 oxidizing dissolution mechanism under irradiation. Thus, for a high H2O2 content, UO2 oxidation leads to the release of U (VI) in solution which may precipitate in the form of studtite (UO44H2O) or, with dehydration, of metastudtite (UO42H2O) [4,10,11]. For less oxidizing conditions, precipitation of the schoepite phase (UO3xH2O) has been observed [1,2]. Concerning MOX fuels and in particular the MIMAS (MIcronized MASter blend) type, whose heterogeneous microstructure is characterized by the presence of Pu-enriched aggregates spread within a matrix mainly containing UO2, little information exists to date regarding their alteration process in underwater storage conditions. Under highly oxidizing conditions (generated by an external gamma irradiation source) and at a neutral pH (6.2–6.9), the release of uranium (U (VI)) and plutonium (Pu (V)) increases significantly compared to a reference experiment without an external irradiation source [12]. The heterogeneous nature of the fuel makes determining the origin of the species released difficult (i.e. coming from the mainly UO2 matrix or the Pu-enriched aggregates). Raman spectroscopy has nevertheless shown the UO2 matrix’s higher sensitivity to oxidation. Fuel matrix oxidation depends significantly on the Pu concentration, as Pu-enriched phases are more stable over time. This supports the theory of a dominant contribution from the UO2 matrix to the releases in solution [13]. Nevertheless, it must be remembered that in studies undertaken to gain a better understanding of fuel alteration mechanisms under water, the pH parameter must be taken into account. It is known that the slightly acidic pH conditions which are to be expected in storage pools will not have much influence on primary yields (compared to pure water) and therefore on the water radiolysis products [14]. They will however have an effect on the speciation of uranium and plutonium in solution, and therefore on the fuel matrix oxidizing dissolution behavior [15,16]. In this paper, we contribute further elements to the understanding of alteration mechanisms of the heterogeneous MOX fuel matrix under radiolytic oxidation in the water of storage pools. This study’s originality lies in its taking into account of pH conditions (slightly acidic) and irradiation, and in the combination of solution analyses with fuel surface characterization at various stages of the leaching process. This data enabled a description of the steps from fuel matrix dissolution through to the precipitation of secondary phases, with time, in representative conditions of underwater interim storage.

231

2. Experimental 2.1. Characteristics of the MIMAS MOX fuel studied To meet the needs of this study, investigations were carried out on a MIMAS MOX TU2Ò fuel sample with an initial enrichment in Pu/(Pu + U) of 6.6 wt.% irradiated in the Dampierre 2 reactor for 4 cycles from 15 November 1993 to 16 May 1998. Its average burnup was 48.8 GWd/tHM. The segments of clad fuel examined came from cutting up a single fuel rod (M09-34275 from the FXPOEH assembly). Because of the fabrication process implemented (dilution of a mixture of UO2 and PuO2 oxides in a UO2 powder) and the origin of the UO2 powder used, this type of MOX fuel has a heterogeneous microstructure characterized by the presence of 3 distinct phases whose characteristics before irradiation were [17]:  A phase with a low plutonium content (approximately 2.7%) but which contains however 15% of the total plutonium. This phase, mainly composed of UO2 will be called the ‘‘mainly UO2 matrix’’ hereafter.  An aggregate phase with a high plutonium content (approximately 20%) where 38% of the total plutonium accumulates. These plutonium-rich aggregates are randomly spread within the fuel matrix.  A phase with an intermediate plutonium content (approximately 7%) which surrounds the two phases mentioned above and contains 47% of the total Pu. After irradiation, the average plutonium content is about 3.6 wt.%. Plutonium-rich aggregates are still present. They were restructured in the reactor (aggregates in the pellet periphery) and have a local burnup rate approximately 2.5 times higher than the average burnup. Certain aggregates reach 15–18% of plutonium but for most of them, the Pu content is approximately 10%. Half of the Pu is evenly distributed between the mainly UO2 matrix and the aggregates. The phase with an intermediate plutonium composition is difficult to observe after fuel irradiation, and its behavior under water is not studied in this document. The mass fractions (xi) in radionuclides for the MOX fuel studied were calculated with the CESAR code [18]. The inventory of the main radionuclides examined during this work is given in Table 1. 2.2. Sample preparation and leaching protocol Two leaching tests were carried out on segments of clad MOX fuel in order to compare the general matrix alteration under different leaching conditions. Two samples were cut from near the top of the fuel rods: a fuel corrosion sample, consisting of 7.5 mm and 8.5 mm segments so less than the entire pellet height (Fig. 1). The samples were pre-leached for one week in carbonated water (103 M, NaHCO3) in order to remove as much as possible of the oxidized uranium layer that can form on the surface of samples during their storage in a hot cell. It is important to eliminate any contribution from this layer, so that only phenomena linked to the MOX fuel matrix alteration under water radiolysis are taken into consideration. After the carbonated water wash step, the leaching experiments in pure aerated water at pH 5–5.5 began:  Exp. 1 (with external c): Experiment carried out in the presence of an external gamma irradiation source. The 60Co source gave a 210 Gy/h dose rate and enabled simulation of the irradiation conditions characteristic of a fuel rod assembly after

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M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241 Table 1 Mass fractions of the main isotopes and the chemical elements in the MOX fuel, calculated using the CESAR code on the date the leaching experiments began (2009). Isotope/element

xi MOX

U Pu Pu 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu 243 Pu 244 Pu

8.01E01 3.47E02 1.76E11 1.35E03 1.30E02 1.22E02 3.81E03 4.33E03 6.75E18 5.17E07

Thermal power

W t1 HM (MOX)

a

3.5E+03 1.3E+03 1.0E+03

b

c

Leaching vessel (stainless steel) Titanium liner (1 mm thickness)

110 mm

Ultrapure water (230 ml) Spent fuel segment (sample holder)

10 mm

236

Air vent

51 mm

(a)

(b)

Fig. 2. Picture (a) and schematic representation (b) of the experimental set-up for the leaching experiments.

30–40 years of cooling. The dose rate from the source was measured, using the Fricke dosimetry method (measure of the ferric ions formed by the oxidation of ferrous ions under the effect of radiation).  Exp. 2 (reference): reference experiment, where the sample was subjected to its own self-irradiation field (around 32 Gy/h gamma rate dose, calculated by the MicroshieldÒ software). In order to be representative of the pH conditions of the water under interim storage conditions (5–5.5), a small content of nitric acid solution (HNO3 0.5 N) was added for this purpose. Because of the experiment were carried out in aerated conditions, this acid addition was realized after that the solution is in equilibrium with CO2 from the air in contact with the solution. These experiments, carried out in a shielded cell, were under a static regime (i.e. with no renewal of the solution, V = 230 ml) at room temperature (25 °C). The vessel for leaching tests is consisting of stainless steel body (Fig. 2(a)) and titanium liner presenting a TiO2 layer (chemically inert) (Fig. 2(b)). The segments were put in contact with the leaching solution thanks to a titanium sample holder placed at 10 mm from the bottom of the titanium liner. Both faces of the segments were in contact with the solution.

A static alteration mode was chosen in order to favor oxidation and the precipitation of secondary phases on the fuel surface, while avoiding any removal of matter likely to have an impact by the large-scale diffusion process or in the presence of an alteration solution flux. The tests lasted 222 days for Exp. 1 and 604 days for Exp. 2. Samples of the solution were taken at different times. Following the tests, samples were filtered (0.45 lm syringe filter, Sartorius) and ultra-filtered by centrifugation (10 000 Da, i.e. a cutoff at 1.8 nm), in order to evaluate the colloids which could have formed in the solution. 2.3. Chemical and radiochemical analysis methods of the solution Elementary and radiochemical analyses of the solution were carried out on samples taken after different periods of time. Uranium was analyzed in solution by KPA (Kinetic Phosphorescence Analyzer). This technique enables uranium analysis in a concentration range from 0.1 to 100 lg/l (4.2  1010 to 4.2  107 mol/l), with a quantification limit of 0.1 lg/l. The different plutonium isotopes were analyzed by alpha

MOX spent fuel

Pre-leaching of MOX fuel clad segment under carbonated and aerated static conditions (1 l) L=7.5 mm (m=4.58 g)

L=8.5 mm (m=3.91 g)

Leaching test (static and aerated) in deionized water (230 ml) at pH 5-5.5 and under external γ irradiation

Leaching test (static and aerated) in

Exp. 1 (with external γ)

Exp. 2 (reference)

deionized water (230 ml) at pH 5-5.5

Fig. 1. Summary of leaching experiments performed on spent MOX fuel clad segments.

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spectrometry, and the plutonium concentration in the solution was determined from the isotope concentrations. Hydrogen peroxide (H2O2) was analyzed in the leaching solution by spectrophotometry following the Ghormley method (reaction of the H2O2 molecules with iodide ions) for levels from 4  106 to 2  104 mol/l and by chemiluminescence for lower levels from 1  109 to 6  107 mol/l. Measurement of the solution’s redox potential during each periodic analysis was carried out using an Ag/AgCl electrode (KCl 3 mol/l) with a platinum probe. The electrode voltage with respect to the hydrogen electrode was +207 mV at 25 °C. The pH was also measured for each sampling. The low conductivity of the distilled water made difficult the measurement of the pH. A dedicated electrode (BlueLineÒ – SI Analytics) for the measurement of the pH in a slightly charged solution was used during both experiments. The volume of the leaching solution varied during the experiment with the samples. All of these samples, however, are limited to 10% or 15% of the initial volume in order to not vary significantly the volume of leaching solution. 2.4. Characterization of the fuel surface states by Raman spectroscopy The evolution of the segments’ surface state was followed by Raman spectroscopy, with special attention to the plutonium-enriched aggregates and the mainly UO2 matrix at the different times. This meant that before leaching, one side of each sample segment was polished (mirror polish) and a selective chemical attack of the polished face of the UO2 grains was carried out. The attack solution was a mixture consisting of H2SO4 (technical solution, 110 Volume, 30 wt.%) and H2O2 (technical solution, 2 M). Several attacks were done, with a total solution sample contact time of 105 s. Between the attacks, the samples were rinsed with distilled water before optical microscope examination. This initial characterization was necessary as it enabled an identification of the plutonium-enriched aggregates within the mainly UO2 matrix in the extreme surface, and therefore simplified the location of the different zones during the Raman spectrum acquisition for these long-term experiments. Before each observation, the segment was removed from the reactor and softly dried with a tissue in order to take off the residual water on the surface of the segment faces. Raman microspectrometry measurements were carried out with a Labram confocal Raman spectrometer (Horiba, Japan). Raman backscattering was excited with a 532 nm excitation

wavelength Nd-YAG laser operating at 1.5 mW to avoid oxidation of the UO2 phase surface during the measurements. Individual spectra were obtained using an exposure of 5  60 s over the wavenumber range from 200 to 1500 cm1. The laser was focused onto the sample using 100 uncoated-objective lenses of the optic microscope (Optique PETER) with a lateral resolution about 0.7 lm. The Raman spectrometer was calibrated using a Si single crystal, and the correct shift was maintained for all samples. The nuclearized optical microscope was set up in the shielded cell, whereas the Raman spectrometer and the laser were outside it. Optical fibers transferred the signal between the two zones. 3. Results 3.1. Leaching solution analyses over time – Total releases of main radionuclides into solution 3.1.1. Leaching solution chemistry: Redox potential, [H2O2] and pH Values of redox potential measured in the solution over time for the Exp. 1 samples (under external c field) and Exp. 2 (reference) are compared in Fig. 3(a). The potential measured for the Exp. 1 solution was +567 mVSHE and was slightly higher than the potential measured for Exp. 2, with +537 mVSHE. These potential values were relatively stable throughout the two experiments. Concerning the formation of H2O2, in the presence of the c field emitted by the source its concentration in the solution reached 7.6  105 mol/l after 222 days of leaching (Fig. 3(b) and Table 2). In the absence of the source, the concentrations were lower and stabilized at about 7.9  106 mol/l after 604 days of leaching. The external gamma irradiation of the solution by the source therefore generated more oxidizing species, and so led to slightly more oxidizing leaching conditions than just in the presence of the segments’ self-irradiation field. This can be explained by the fact that the self-irradiation field is essentially local and mainly concerns a (range of 40 lm in water) and b rays (range of 300– 400 lm in water). The concentration profile of the oxidizing species is not homogeneous and decreases outside the zone irradiated by the segment because of the diffusion of species into the only gamma irradiated part of the solution [19,20]. On the contrary, the source generates oxidizing species homogeneously within the solution. For the pH, values measured in the solutions were stable throughout the experimental periods and were around 5–5.5 in

Time (day) 0

700

50

100 150 200 250 300 350 400 450 500 550 600

1E-4

500

Exp. 1 (with external γ) Exp. 2 (reference)

400

[H2O2 ] (mol/l)

EhSHE (mV)

600

1E-5

1E-6

Exp. 1 (with external γ) Exp. 2 (reference)

300

200 0

50

100 150 200 250 300 350 400 450 500 550 600

1E-7

Time (day)

(a)

(b)

Fig. 3. Evolution (a) of redox potential (Ag/AgCl electrode (KCl 3 mol/l)/Pt) and (b) of the hydrogen peroxide concentration (H2O2) in the leaching solutions during Exp. 1 (with external c field) and Exp. 2 (reference).

234 Table 2 Total concentrations in actinides (uranium and plutonium) released in solution and H2O2 concentrations formed during Exp. 1 (with external c field) and Exp. 2 (reference) leaching experiments. Thermodynamic calculations, using CHESS v3.0 and the EQ3/6 database, for uranium and plutonium concentrations at the thermodynamic equilibrium and distribution of uranium and plutonium species in the alteration solution. Calculations for the saturation indexes (SI) of the studtite, schoepite phases. Time (days)

Exp. 2 (reference) 1 7 18 31 59 89 176 365 604 direct sample 604 filtered sample 604 ultra-filtered sample

[H2O2] (mol/l)

pH

[Utot] exp. (mol/l)

[UO2+ 2 ]eq (mol/l)

[UO2OH+]eq (mol/l)

[UO2(OH)2aq]eq (mol/l)

– 577 567 567 567 577

– 3.6  105 3.9  105 3.6  105 4.6  105 7.6  105

5 5 5 5 5.5 5.5

3.5  107 1.6  106 3.3  106 2.6  106 1.5  106 2.9  107 2.0  106 8.4  108
3.8  109 3.6  109 3.9  109 2.9  1010 1.7  1010 1.8  1010 1.7  1010 –

2.3  109 2.2  109 2.4  109 5.7  1010 3.3  1010 3.4  1010 3.3  1010 –

1.8  109 1.7  109 1.9  109 1.4  109 8.3  1010 8.4  1010 8.2  1010 –

557 548 549 537 527 532 537 527 527

– 3.5  107 5.0  107 8.8  107 3.9  106 5.6  106 4.2  106 7.6  106 7.9  106

5 5 5 5 5 5 5.5 5.4 5.4

5.3  108 1.8  107 4.3  107 7.9  107 2.9  106 8.8  106 6.4  105 8.3  106 1.4  105 4.0  108 3.4  108

8.6  108 2.0  107 2.3  107 1.1  107 1.5  106 6.8  107 1.5  107 9.3  107 2.6  109 2.6  109

5.3  108 1.3  107 1.4  107 6.5  108 9.2  107 1.3  106 2.3  107 1.4  106 4.0  109 4.0  109

4.1  108 9.8  108 1.1  107 5.1  108 7.2  107 3.2  106 4.5  107 2.8  106 7.9  109 7.9  109

SI = 0: solubility limit of the secondary phase, SI < 0: under-saturation and SI > 0: over-saturation. NC: not calculated because of no redox potential measurement. QL: quantification limit. a

b

SIstudtite ¼ log

½UO2þ ½H2 O2  2 2 ½Hþ 

K Sstudtite

.

½UO2þ  2 þ 2

SIschoepite ¼ log K S ½H

schoepite

.

[Utot]eq. (mol/l)

SI (studite)a

SI (schoepite)b

[Putot] (mol/l)

[PuO+2] (mol/l)

[Pu(OH)4aq] (mol/l)

– 7.5  109 6.9  109 7.5  109 2.1  109 1.3  109

– 2.3 2.6 2.5 2.8 2.3 3.2 1.8 –

– – – – – – – – –

1.1  108 3  108 7.1  108 8.5  108 2.1  107 3.2  107 1.3  107 2.3  107 1.7  107

– 8.6  109 5.8  109 5.8  109 7.4  109 8.6  109

– 1.7  109 1.7  109 1.7  109 1.7  109 1.7  109

– 3.2  107 5.4  107 3.1  107 7.0  108 4.8  108 2.4  108 1.5  108 1.4  108

– <0 <0 0.4 1.6 2.2 2.9 2.6 2.8 0.4 0.3

– – – – – – 0.5 – – – –

1.2  109 1.2  108 1.3  108 4.7  108 3.1  107 2.5  107 3.8  107 8.6  107 1.3  106 2.8  107 6.8  107

7.9  1010 2.8  109 2.9  109 1.8  109 1.2  109 1.5  109 1.8  109 1.2  109 1.2  109

3.4  1010 1.7  109 1.7  109 1.7  109 1.7  109 1.7  109 1.7  109 1.7  109 1.7  109

M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241

Exp. 1 (with external c) 1 7 20 28 120 222 direct sample 222 agitated sample 222 filtered sample 222 ultra-filtered sample

Eh (mVSHE)

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M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241

Time (day) 0

50

100 150 200 250 300 350 400 450 500 550 600

Exp. 1 (with external γ) Exp. 2 (reference)

1E-4 Direct sampling

[U] (mol/l)

1E-5

Solubility limit of schoepite Direct sampling

Agitated sampling Filtered sampling

1E-6 Direct sampling

1E-7

Filtered sampling

Filtrered sampling Ultra-filtered sampling Solubility limit of studtite without γ irradiation

1E-8 Solubility limit of studtite underγ irradiation

1E-9

Fig. 4. Evolution of uranium concentrations (in mol/l) analyzed in the leaching solution during the Exp. 1 (with external c) and Exp. 2 (reference). The uranium solubility limits compared to the precipitation of the studtite and the schoepite phases, calculated depending on the conditions of pH and H2O2 concentration of the experiments, are also reported on the graph (dotted lines). Error bars – two standard deviations (less than plotted data points).

both cases. The pH interval 5–5.5 was due to the spread of the measured pH values and it had not evolved during time.

3.1.2. Uranium release into the leaching solution The uranium concentrations released into the solution for the two experiments are shown in Fig. 4 and Table 2. With the external c field (Exp. 1), the concentrations increased quickly up to 3.3  106 mol/l after 20 days of leaching, and then stabilized (2.6  106 mol/l) or even decreased (2.9  107 mol/l). The agitation of the solution realized at the end of the experiment enabled a U concentration of 2.0  106 mol/l to be found, close to that measured after 20 days (3.3  106 mol/l). Without the external c field, U concentrations increased up to 6.4  105 mol/l after 176 days and then remained stationary at around 1.4  105 mol/l after 604 days of leaching. The same trends for U release into the solution could therefore be seen for the two experiments, with a stationary concentration state reached. The difference between the experiments came from the release kinetics, which were slower for the experiment without the external c field and therefore reached U concentration stability later. After 50 days of leaching, the U releases into the solution for this experiment were higher than those measured for the samples obtained under gamma field.

The filtration and ultra-filtration of the solutions, carried out at the end of the two experiments, led to a high U loss of up to three orders of magnitude for the ultra-filtered solutions (Fig. 5 and Table 2). Therefore, a large part of the released uranium is in the form of a precipitate in the solution. This would explain why the U concentration measured after agitation of the solution was higher than that of the Exp. 1 solution, sampled directly after the opening of the reactor (without agitation). It should be noted that for Exp. 1 after 222 days of leaching, the U concentrations could not be measured in the ultra-filtered solution. It can be supposed that the U concentration in this ultra-filtered solution was close to 1  109 mol/l, corresponding to the quantification limit (QL) for U by KPA.

3.1.3. Plutonium release into the solution Concerning plutonium, during the experiment under external c field (Exp. 1), the concentrations analyzed in the solution were relatively high (3.2  107 mol/l after 222 days of leaching) and comparable to those obtained for the reference experiment (Exp. 2) (1.3  106 mol/l after 604 days) (Fig. 6 and Table 2). In both cases, the Pu releases increased progressively over time and remained relatively constant throughout the experiments (no concentration stabilization). Unlike the case of uranium, filtration and ultra-filtration of the leach solutions for the two experiments did not lead to any significant variation of the plutonium concentration (considering measurement uncertainties). This would tend to indicate that there were no Pu-based colloids (of a size greater than the cutoff of the ultrafiltration, which was 1.8 nm) in the leach solutions from the two experiments.

3.2. Investigation of the fuel matrix surface state 3.2.1. Initial fuel matrix characterizations (before leaching) The selective acid attack (H2SO4/H2O2) on the UO2 grains carried out on the fuel segment surfaces before their positioning under water enabled visual detection of the plutonium-enriched aggregates from the rest of the mainly UO2 matrix (Fig. 7(a)). Before starting the experiments, the segments were pre-leached in carbonated water for one week. After this pre-leaching, the surface state of the fuel matrix was examined by Raman spectroscopy, in order to characterize it at time t = 0 and be able to follow changes as the experiments progressed. The mainly UO2 matrix and the Pu-enriched aggregates were characterized distinctly (Fig. 7(b)) for the two fuel segments.

(a) Exp. 1 (with external γ ) after 222 days of leaching

-6

(b) Exp. 2 (without γ ) after 604 days of leaching

10 -5

10 -6

10-7

[U] mol/l

[U] mol/l

10

10 -7

10-8


10-9

Sample

Filtered sample (<0.45 µm)

[U]eq

Ultra-filtered sample (<1.8 nm)

[U]eq 10 -8

Sample

Filtered sample (<0.45 µm)

Ultra-filtered sample (<1.8 nm)

Fig. 5. Evolution of the uranium concentrations (mol/l) in the leaching solution after filtration (0.45 lm) and ultra-filtration (1.8 nm), at the end of the experiments (a) Exp. 1 (with external c) and (b) Exp. 2 (reference), i.e. after 222 days and 604 days of leaching respectively. The concentrations at the thermodynamic equilibrium, calculated using the JCHESS calculation code and the EQ3/6 thermodynamic base, are shown by the dotted line.

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M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241

Time (day) 0

100

200

300

[Pu] (mol/l)

1E-8

500

600

Direct sampling

1E-6

1E-7

400

Direct sampling Filtered sampling

Ultra-filtered sampling

Direct sampling Filtred sampling Ultra-filtered sampling Agitated sampling

Filtered sampling

Exp. 1 (with external γ ) EXp. 2 (reference)

1E-9

Fig. 6. Evolution of plutonium concentrations (mol/l) in the leaching solution during the two experiments: Exp. 1 (with external c) and Exp. 2 (reference). Error bars – two standard deviations.

3.2.1.1. Initial surface state of the mainly UO2 matrix. The Raman spectra of this matrix showed the presence of three main bands (445, 576, 1150 cm1) (Fig. 7(b)). These bands are characteristic of the UO2 phase which has a fluorite-type structure of the Fm3m space group (Oh symmetry) [21]. Group theory predicts six optic phonon branches, a triply degenerate Raman active mode T2g and an infrared (IR) active mode T1u for fluorite structure. The corresponding T2g vibrational mode appears at around 445 cm1. The presence of defects distorts the translational symmetry and relaxes the selection rules, making several normally dipole-forbidden optical transitions observable. The intense feature observed in the Raman spectra around 576 cm1, assigned to be the IR active T1u longitudinal optic (1LO) mode, appears to be due to the breakdown of selection rules as a result of presence of defects [22,23]. A broad band observed around 1150 cm1, a band initially assigned to an electronic crystal field transition, is now attributed to the second-order longitudinal optic (2LO) phonon based on the similarity of 1150 cm1 mode with 576 cm1 in terms of resonance profile, pressure dependence intensity, and frequency [24]. The origin of the 2LO band is still the subject of debate in the literature [24,25], but it remains the indicator of the amount of UO2 oxidation. In particular, its intensity may decrease for low differences in stoichiometry [25]. A shoulder can also be seen at 627 cm1. Its origin has been attributed to anionic sub-network distortions during UO2

oxidation into UO2+x by He and Shoesmith [26], and more specifically to a vibration mode involving a cuboctahedric-type structure (U4O9) by Desgranges et al. [27]. As well as these 4 contributions (445, 576, 627 and 1150 cm1), the Raman spectrum fitting of the mainly UO2 matrix in the frequency range of 400–700 cm1 indicates the presence of a band located at around 530 cm1. This band has already been noticed for fission product-doped UO2 (SIMFUEL) [28], for Nd-doped UO2 (U,Nd)O2x-type) [29] and recently for samples of UO2 leached in pure water under alpha irradiation [30]. It is attributed to sub-stoichiometric defects with the incorporation of oxygen vacancy in the UO2 lattice as (U,Nd)O2x-type by Desgranges et al. [29]. Elsewhere, Guimbretière et al. [30], showed that its presence is linked to the effects of UO2 phase irradiation and that it evaluates with the 627 cm1 band (footprint of a UO2+x-type hyperstoichiometric phase). Thus even if the origin of the 530 cm1 band is not yet completely established, its contribution is essential during the deconvolution of Raman spectra acquired on the mainly UO2 matrices of MIMAS-type MOX fuels. Based on the works of Guimbretière et al. [30], the bands centered around 530 cm1, 576 cm1 (1LO) and 627 cm1 (named ‘‘triplet of defect bands’’ by Guimbretiere et al.) will hereafter be referred to as U1, U2 and U3 respectively. Further study of the U3 and 2LO bands contributes information enabling precision regarding the fuel matrix oxidation state at the initial stage (Fig. 7(b)). The 2LO band’s intensity indicates that the mainly UO2 matrix is relatively stoichiometric. However, the contribution from the U3 band affects this fact, pointing out that there are hyperstoichiometric phases on the mainly UO2 matrix surface. The carbonated water wash step which enables uranium (VI) to be complexed by carbonate ions did not give a complete removal of these UO2+x phases. The absence of any U3O8-type oxidized phase and studtite- or schoepite-type secondary phases on the sample surfaces were also checked before their leaching. For the initial stage, no difference was observed in the Raman spectra of one segment compared to another. Only the Raman spectra for the sample used in Exp. 1 (under external gamma irradiation) are therefore given in Fig. 7(b). 3.2.1.2. Surface state of the Pu-enriched aggregates. The Raman spectra recorded on the surface of the Pu-enriched aggregates before the leaching are given in Fig. 7(b). It should be remembered that the average Pu concentration in the aggregates was 10–15% after fuel irradiation. The PuO2 phase have the same cubic fluorite-type

576

1.5 576

445

UO2 matrix

Pu-aggregate x

445

627 530

Normalized intensity

x

627

1.0

576 400

456

UO2 matrix

600

576

0.5 456

635

Pu-aggregate 400

20 µm

600

0.5 200

400

600

800

1000

1200

1400

Raman shift (cm-1)

(a)

(b)

Fig. 7. (a) Optical picture of the different zones analyzed on the surface of the fuel matrix by Raman spectroscopy (mainly UO2 matrix and Pu-enriched aggregates). (b) Raman spectra for the mainly UO2 matrix and for a Pu-enriched aggregate, carried out after the sample pre-leaching step, just before the leaching experiments began.

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Exp. 1 (with external γ ) 576

2.5

445

1150

627 581

1.5

Pu-aggregate

1.0

Initially

Normalized intensity

UO2 matrix

456

2.0

819

0.5 865

UO4.4H2O After 20 days of leaching

20 µm

0 200

400

600

800

1000

1200

1400

Raman shift (cm-1)

(a)

(b)

Fig. 8. (a) Optical picture of the fuel matrix surface observed after 20 days of leaching under Exp. 1 conditions (with external c). (b) Evolution of Raman spectra observed on the fuel matrix surface over time. Initially: the Raman spectra of UO2 matrix and Pu-aggregate were specifically observed. After 20 days of leaching: only the Raman spectrum of the studtite phase (UO44H2O) was recorded.

structure as the UO2 phase. Thus, the 456 cm1 band seen on the Raman spectra for the aggregates is associated with the fluorite structure T2g vibration mode of the solid solution (U,Pu)O2 (Fig. 7(b)). In the case of Pu aggregates, the vibration frequency for this band is higher than that of UO2 (445 cm1). The work of Jégou et al. [13] showed that the position of the T2g scattering active in Raman for these fluorite structures (Oh symmetry) depends on the plutonium content. While it is at around 445 cm1 in the case of UO2, it tends to move toward higher wavenumbers when the plutonium content is higher in the solid solution. It appears at 457 cm1 for UO2 pellets doped with 25 wt.%. of Pu (Pu content close to that present in Pu-enriched aggregates found in MIMAS-type MOX fuels before their irradiation) and evolves up to 463 cm1 or even 470 cm1 for samples mainly containing plutonium (239PuO2 powder) [13]. According to the authors, this can be interpreted by a PuAO bonding strength higher than that of UAO, giving a shorter PuAO bond length in the fluorite structure and therefore an increase in the vibration frequency of this bond. In addition to the T2g band, the U2 (576 cm1) and U3 (627 cm1) bands can also be seen in the aggregates’ Raman spectra. The contribution from the 627 cm1 (U3) band linked to the hyperstoichiometry of the UO2 phase is less intense on the Raman spectra acquired on the surfaces of the Pu-enriched aggregates than on those of the mainly UO2 matrix (Fig. 7(b)). This difference could be explained, on the one hand, by a lower UO2 content in the aggregates than in the mainly UO2 matrix, and consequently a smaller contribution from the UO2+x hyperstoichiometric phase. On the other hand, the presence of Pu could have an effect leading to an oxidation limitation for the UO2 phase present in the aggregates [13]. The 2LO (1150 cm1) resonance band is not seen on the Pu-enriched aggregate Raman spectra. This phenomenon is probably explained by the fact that the Raman spectrometer excitation source energy (2.34 eV for a 532 nm wave-length laser) is so far from the band-gap energy of (U,Pu)O2 solid solution, estimated at around 1.1–1.5 eV by Dorado and Garcia [31] for MOX fuel containing 13% of plutonium. A final point that must be emphasized concerns the U1 band of the defect triplet (530 cm1). This band, associated with

sub-stoichiometry defects [29] or with irradiation effects [30], is not taken into account in the fitting of the Pu aggregates Raman spectra. Its contribution is negligible (Fig. 7(b)). The presence of Pu at high concentrations in the aggregates (10–15%) seems to decrease the contribution from the structural defect observed for the mainly UO2 matrix (oxidation or irradiation effect). 3.2.2. Fuel matrix surface state over the leaching experiment time 3.2.2.1. Leaching condition: under external gamma field. In the experiment carried out under an external c field, a superficial layer of crystals was noted over the entire segment surface after only 20 days of leaching, as shown in the optical picture in Fig. 8(a). The Raman spectrum of these crystals features fine bands centered at 819 cm1 and at 865 cm1 which are attributed, respectively, to the symmetric stretching vibration of the U@O bond (UO2+ 2 ) and to the vibration of the OAO bond (O2 2 ) in the structure of the studtite phase (UO44H2O) (Fig. 8(b)) [10]. Thus during this experiment, studtite phase precipitation was observed, occurring extensively (complete covering of the fuel matrix) and rapidly (after only 20 days of leaching). The microstructure of the studtite phase formed in these leaching conditions is in needle-form (Fig. 8(a)). The presence of this phase on the fuel matrix surface meant it was not possible to specifically study the oxidation state of the mainly UO2 matrix and the Pu-enriched aggregates over time. 3.2.2.2. Leaching condition: without external gamma field. For the experiment without the external c field, the Raman spectra for the mainly UO2 matrix acquired over time, and more particularly the 5 bands (T2g, U1, U2 U3 and 2LO) indicate that there was no, or very small, evolution of their contribution for up to 176 days of leaching (Fig. 9(a)). The same trend was the same for the aggregates, as there is no significant change in the contribution from the different Raman bands. The 3 bands (T2g, U2 and U3) are always present with shifts and intensities similar to the initial spectra. The fuel matrix (mainly UO2 matrix and Pu-enriched aggregates) oxidation state does not change significantly over time for 176 days of leaching, according to the Raman spectroscopy characterization results. After 176 days however, bands very different from that of the mainly UO2 matrix or of the Pu-enriched aggregates, with the presence of main bands at 838 cm1 and 852 cm1, was seen at the surface of the mainly

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Exp. 2 (reference ) 10 576 445

9

627

1150

579

UO2 matrix

576

Pu-aggregate

456

8

Initially

7 444

Normalized intensity

1150

627

532

6 457

580

UO2 matrix

5

Pu-aggregate 4

576 444

After 176 days of leaching

1150 627

3

838

UO2 matrix + UO3.2H2O

852 246 886

2

20 µm

819

(b) 1

864

UO4.4H2O 0 200

400

600

800

1000

1200

After 365 days of leaching

1400

Raman shift (cm-1)

(a) Fig. 9. (a) Raman spectra recorded on the fuel matrix surface over time under Exp. 2 conditions (reference). The schoepite phase (UO32H2O) bands were observed after 176 days of leaching on the spectra of the mainly UO2 matrix. After 365 days, only the Raman spectra for the studtite phase (UO44H2O) were observed on the sample surface. (b) Optical picture of the fuel matrix surface observed after 365 days of leaching under Exp. 2 conditions.

-3

-3

(a) Exp. 1 (with external γ ) [H2O2]=7.6x10 -5 mol/l - 222 days

Log(Uranium total) (mol/l)

-4

UO2 (OH) 42-

-5

UO4.4H2 O (studtite)

-6

UO2+ 2 -7

UO2(OH) -3

Log(Uranium total) (mol/l)

(b) Exp. 2 (reference) [H2O2]=7.9x10 -6 mol/l - 604 days -4

UO2 (OH) 24

-5

UO4.4H2O (studtite)

-6

UO2+ 2 UO2(OH) -3

-7

-8

-8

-9

-9

UO2(OH) 2 (aq) 0

2

4

5

6

8

10

12

14

pH

0

2

4

5

6

8

10

12

14

pH

Fig. 10. Predominance diagram of uranium species depending on pH established for (a) Exp. 1 (with external c field) with [H2O2] = 7.6  105 mol/l (concentration obtained at the end of the experiment, after 222 days of leaching) and (b) Exp. 2 (reference) with [H2O2] = 7.9  106 mol/l (concentration obtained at the end of the experiment, after 604 days of leaching).

UO2 matrix. These bands were attributed to the schoepite secondary phase. Very few spectra for this phase were noted during the Raman characterizations carried out during the 176-day observation, nor were any others found during other periodic observations. Later, for other periodic observations, after 365 days of leaching, specific analyses of the mainly UO2 matrix became impossible due to the formation of a superficial layer of crystals, identified as studtite

(UO44H2O). This phase covered the whole surface of the fuel matrix, as shown by the optical picture in Fig. 9(b). During this experiment, therefore, studtite precipitation was noted, with a kinetically slow formation (after 365 days of leaching) but widespread, with total coverage of the fuel matrix surface. The studtite microstructure observed under the Exp. 2 (reference) conditions had a structured geometry with the formation of polyhedrons (Fig. 9(b)).

M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241

4. Discussion 4.1. Understanding the release mechanisms for uranium in solution leading to the precipitation of secondary phases 4.1.1. Studtite phase precipitation For both experiments (with and without the external c field), a phenomenon of uranium saturation in the solution with a stable concentration level reached was recorded during the fuel matrix alteration. The time to reach this stable level depended on the solution irradiation conditions: it is faster with the external c field, taking just 20 days compared to 176 days of leaching without this field. To explain this trend toward stability of U concentrations in the solution, two hypotheses can be examined involving the equilibrium between the surface and the solution: (i) appearance of a precipitation phenomenon with a uranium-based secondary phase, and in this case the uranium concentration in the alteration solution depends on the solubility of this phase, or (ii) fuel matrix oxidation with the formation of an oxidized phase of the U3O7 or U4O9 type, leading to a new equilibrium with the solution. Optical observation of the samples during their alteration enabled the identification of the phenomenon causing this saturation with the presence of secondary phases on the fuel matrix surface. Raman spectroscopy indicated that these phases were of studtite- (in the majority) or schoepite-type (to a lesser extent). These phases were found on the fuel matrix surface as soon as the U concentrations reached a saturation state in the solution (which is not the thermodynamic equilibrium). The good correlation of the information obtained by solution analyses and from fuel matrix characterization can be noted. Moreover, the absence of any change in the 2LO (1150 cm1) and U3 (630 cm1) bands in terms of wavenumber and spectral intensity shifts showed that the fuel matrix oxidation state (mainly UO2 matrix and Pu-enriched aggregates) did not change significantly during the fuel alteration for the experiment without the external c field, compared to the initial state. Studying the fuel matrix oxidation state in the conditions of the experiment with the c field was made impossible by the massive presence of the studtite phase found as early as the first observation. Through the JCHESS calculation code [32] and the thermodynamic basic data of the EQ3/6 code [33], it is possible to establish the predominance diagrams of uranium species for the leaching conditions of the two experiments (Fig. 10). The diagrams obtained for Exp. 1 (under external c field) with [H2O2] = 7.6  105 mol/l (Fig. 10(a)) and for Exp. 2 (reference) with [H2O2] = 7.9  106 mol/l (Fig. 10(b)) indicate that the alteration conditions are favorable to the formation of studtite phase. The thermodynamic calculations, made with the same codes, also enabled determination of the concentrations in U-base species formed during the matrix alteration. These concentrations evolved + following [UO2+ 2 ] > [UO2OH ] > [UO2(OH)2aq] (where the uranium is in VI oxidation state) (Table 2). The majority uranium species in these leaching conditions is therefore the uranyl ion UO2+ 2 . It was shown that in the presence of hydrogen peroxide UO2+ 2 reacts to form the studtite phase (UO44H2O), whose formation reaction is generally defined by: þ UO2þ 2 þ H2 O2 þ 4H2 O () UO4 :4H2 OðsÞ þ 2H ½11; 34; 35

ð1Þ

And the solubility constant corresponding to this reaction at ambient temperature is Ks = 1.37  103 [35]. The uranyl ion is the uranium-based species directly involved in the studtite phase formation reaction. From the concentrations of uranium and H2O2 measured in the solution, and taking the pH conditions into account, it is possible to calculate the U concentration at thermodynamic equilibrium in the

239

solution, i.e. the U concentration at the studtite phase solubility limit ([Utot]eq in Table 2; black and blue1 dash line in Figs. 4 and 5 for both experiments). It was evaluated as 1.3  109 mol/l for Exp. 1 (under external c field) and at 1.4  108 mol/l for Exp. 2 (reference). The ultra-filtration of the solution, carried out at the end of the experiments, enabled an approximate assessment of the experimental concentration values, with 3.4  108 mol/l for Exp. 2. For Exp. 1 the uranium value seemed to be very near the quantification limit by KPA (1.0  109 mol/l) and therefore it could not be measured (Table 2). Here the good correlation between the experimental results and the associated thermodynamic data could be mentioned. The uranium concentrations measured in the direct sampling solutions from the two experiments were higher by 2 or 3 orders of magnitude than the studtite phase solubility limit, under the conditions of H2O2 concentration and pH present in the solution, during all the periodic analyses (except for the first analyses of Exp. 2). If the saturation index (SI) is calculated for the solution compared to the formation of the studtite phase according to Eq. (2): ½UO2þ ½H2 O2  2 2

SIstudtite ¼ log

½Hþ 

K sstudtite

;

ð2Þ

it is shown that studtite precipitation could be observed after 31–59 days of leaching for Exp. 2 (with SIstudtite > 0, Table 2). However, this crystallization phenomenon, involving the stability of the U concentration in the solution (Fig. 4), was not identified by Raman spectroscopy after 365 days of leaching (Fig. 9). There was therefore a delay in studtite precipitation under these alteration conditions with only the presence of the fuel segment self-irradiation field. An over-saturation of the alteration solution in uranium is therefore necessary before studtite precipitation can be observed. In the literature, the description of the mechanisms governing the formation of precipitates in aqueous solution have noted 4 distinct steps: generation of the charge-free precursor [U(OH)Z(OH2)N-Z]0, germination and growth of germs, and ageing of the particles in suspension [36]. Thus, the need to over-saturate the solution in uranium before being able to see studtite precipitation indicates that it is the first steps of these mechanisms that impose limits. The generation of the charge-free precursor able to condense and form the crystalline phase and the condensation of these precursors delay the appearance of studtite without, in the end, changing the thermodynamics over the long term. A minimal concentration (critical threshold) in uranium is then necessary for the formation of germs. Here, for the experiment under external c field, the over-saturation conditions of the solution were reached as soon as the first periodic analyses (Table 2). Thus, the irradiation conditions generated by the c source favor the precipitation of this phase by influencing the mechanisms of germination/growth involving the studtite phase precipitation. The precipitation kinetics of this phase are therefore faster in these leaching conditions and appear after just 20 days under water, as compared to 365 days with only the influence of the segment’s self-irradiation field. 4.1.2. Schoepite phase precipitation During the Exp. 2 (and only during this experiment), the formation of another secondary phase was identified: schoepite. This phase was only noted during the 176-day analyses. Its appearance involved the following reactional equilibrium (with the solubility constant Ks = 0.483): þ UO2þ 2 þ 3H2 O () UO3  2H2 OðsÞ þ 2H ;

ð3Þ

1 For interpretation of color in Figs. 4 and 5, the reader is referred to the web version of this article.

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Fig. 11. Recapitulative scheme of the oxidizing dissolution mechanisms in aerated pure water and at slightly acidic pH (5–5.5) for MOX fuels with (a) the initial state of the spent fuel matrix and (b) the state during the radiolytic dissolution in these leaching conditions.

with ½UO2þ  2 2

SIschoepite ¼ log

½Hþ 

K sschoepite

ð4Þ

The calculations of the solution saturation index for this phase (Eq. (4)), calculated at this periodic interval, indicate that the U releases have been such as the solubility limit of this phase is reached and SIschoepite > 0 (Table 2). In the presence of H2O2, the dominant precipitated phase is studtite [4,10,11] but the need to over-saturate the solution in U delays its precipitation under the Exp. 2 conditions (i.e. without external c field). Uranium therefore continues to be released into the solution in the form of U (VI). As the U concentrations increase, the schoepite phase solubility limit is reached and it can therefore be observed after 176 days of leaching without an external c irradiation field. As soon as the studtite phase germination/growth process begins, the schoepite phase is no longer observed. Thus even if the schoepite phase has been observed, over the long term studtite remains the dominant secondary phase under the leaching conditions studied during this work. 4.2. Plutonium release and chemistry in the alteration solution As the redox potential conditions present during the 2 leaching experiments were relatively similar (537–567 mVSHE), the behavior of plutonium during the fuel matrix alteration under water is comparable from one experiment to the other with continuous releases through to the end of the experiments. The geochemical calculations (JCHESS – EQ3/6) indicate that the two species formed in the solution are mainly: PuO+2 (V), which is the dominant species, and Pu(OH)4aq (IV). These results are correlated with the Pourbaix diagrams established at equivalent potential given in the literature [37,38].

When the solid PuO2 phase was in contact with water, the Pu (IV) ions was released in solution and was oxidized by the very oxidative solution conditions. This should be the main mechanism causing the increase of Pu concentrations. Moreover, analyses of the filtered and ultra-filtered solutions carried out at the end of each experiment indicated that there was a low plutonium-based colloid presence in the solutions. The formation of amorphous phase as Pu(OH)4am was therefore not involved. 4.3. Comparative study of the behavior of Pu-enriched aggregates and mainly UO2 matrix In leaching experiments carried out by Jégou et al. [12] on a segment of MOX fuel irradiated at 47 GWd/t with alteration conditions similar to those described here, (pure aerated water, in the presence of an external c irradiation source), but at pH 6.2–6.9, the dominant uranium species in the solution was UO2(OH)2(aq). From a thermodynamic point of view, this species is not directly involved in the studtite phase formation reaction [35]. The precipitation of this phase, which depends on the UO2+ 2 species present in a very low proportion under these pH conditions, is therefore limited. Moreover, the formation of the UO2+x or even U3O8 phase was seen on the fuel matrix surface. Thus, in this case, the stability of the U concentrations observed as soon as the first days of leaching (under an external c field) depends on the formation/dissolution equilibrium of this phase in the alteration solution. As this phase formed only on the surface of the mainly UO2 matrix, the authors concluded that the Pu-enriched aggregates have better stability as regards oxidizing dissolution. For the experiments described here, the comparative study of the behavior of the mainly UO2 matrix and the enriched aggregates as concerns oxidizing dissolution proved to be rather complex. Under external c irradiation, the fuel matrix surface was covered by the precipitated studtite layer too quickly to be able to carry

M. Magnin et al. / Journal of Nuclear Materials 462 (2015) 230–241

out a comparison. On the other hand, for the experiment without an external gamma field, it was possible to compare the behavior of the aggregates and the mainly UO2 matrix over a period of 176 days. It was noted that there were no major structural modifications of the two phases during this period. It is therefore not possible to make a conclusion as to any better stability of the Puenriched aggregates regarding oxidizing dissolution, under the alteration conditions studied in this paper (pure aerated water and pH 5–5.5). 5. Conclusion The behavior study of the MOX fuel matrix undergoing radiolytic oxidation in aerated conditions at pH 5–5.5 was carried out by coupling chemical and radiochemical analyses of the alteration solution with spectroscopic characterizations of the fuel matrix surface state. The data acquired during the experiments described in this paper enabled the establishment of the reaction mechanism described in Fig. 11. It has been demonstrated that at a slightly acidic pH (5–5.5), the dominant uranium species in solution is UO2+ 2 (VI) and the release of uranium into the solution is controlled by the solubility of the studtite secondary phase over the long term. There is a massive precipitation of this phase, and the kinetics depends on the irradiation conditions. A uranium over-saturation of the solution (minimal concentration) is necessary to observe the precipitation of this phase. If these over-saturation conditions are not reached, precipitation of the schoepite phase can be observed, but in a more limited way compared to the extent of the studtite precipitation phenomenon. As soon as the studtite phase germination/growth process begins, schoepite is no longer observed. Under these pH conditions, thermodynamic calculations indicate that the formation of over-stoichiometric UO2+x-type phases does not intervene in the fuel matrix radiolytic dissolution mechanism. This was correlated by the Raman spectroscopy results, which show that the fluorite structure of the UO2 phase is not impacted by any major structural modifications that could be generated by an evolution of the stoichiometric phase. As concerns plutonium, its release into the solution remains constant throughout the experiments. The dominant plutonium species is PuO+2 (V) and the formation of Pu-based colloid ((PuOH)4am) is not involved. The Pu-enriched aggregates are also unaffected by any change in their structure during radiolytic alteration. Under these pH conditions (5–5.5), spectroscopic analysis of the sample surfaces does not show any effect of the presence of a high Pu content (10–15%) aggregates and does not enable any conclusion as to a better stability of the aggregates compared to the mainly UO2 matrix concerning oxidizing dissolution. The heterogeneous microstructure characteristic of MIMAStype MOX fuels makes the interpretation of analysis data difficult, and therefore the identification of the origin of the nuclides released in solution during the underwater alteration. It is therefore foreseen to carry out leaching experiments in underwater storage conditions (aerated medium, pH 5–5.5 and under gamma irradiation) with fuels presenting a homogeneous microstructure. Leaching a solid solution pellet (U,Pu)O2 with proportions of uranium and plutonium representative of those present in the aggregates of MIMAS-type MOX fuels is also under consideration. Moreover, it seems important to obtain data on the MOX fuel matrix alteration rates under these conditions. The use of fuel samples as segments does not allow the removal of labile inventories present in the grain boundaries, and therefore the use of fission products (134/137Cs and 90Sr) as fuel matrix alteration tracers. Leaching experiments on pre-leached fuel fragments will be

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