Thermal- and radiation-enhanced diffusion of uranium in oxidised zirconium

Surface & Coatings Technology 196 (2005) 10 – 14 www.elsevier.com/locate/surfcoat

Thermal- and radiation-enhanced diffusion of uranium in oxidised zirconium N. Be´rerda, N. Moncoffrea,*, A. Chevariera, Y. Pipona, H. Faustb, H. Catalettec a

Institut de Physique Nucle´aire de Lyon, 4, rue Enrico Fermi, 69622 Villeurbanne cedex, France b Institut Laue Langevin, BP 156, 38042 Grenoble cedex 9, France c EDF R&D, Site des renardie`res, 77818 Moret sur Loingt, Cedex, France Available online 10 February 2005

Abstract This paper is devoted to the study of the defect influence on uranium diffusion in zirconia in the context of nuclear waste disposal. The experiments in reactor conditions are performed at the Institut Laue Langevin in Grenoble using the Lohengrin spectrometer. A thin UO2 layer in direct contact with a zirconium foil is irradiated in the ILL high-flux reactor. The fission product rate is around 3
1011 ions s1 and the neutron flux is equal to 5
1014 n cm2 s1. The target temperature is controlled by an IR pyrometer and ranges from 470 to 490 8C. In these conditions, a zirconium oxidation is first observed before uranium diffusion. A model is proposed to deduce an apparent uranium diffusion coefficient in zirconia (ZrO2) from the energy distribution broadening of a selected fission product (A=90). It is found to be equal to 1015 cm2 s1. The study of thermal diffusion is performed by using ion beam techniques (ion implantation and Rutherford Backscattering). ZrO2 samples are implanted with 800 keV uranium ions at a dose of 1016 ions cm2 and annealed at a pressure of 7.5
101 Pa. No uranium diffusion could be observed up to 800 8C. The influence of irradiation defects mainly due to fission products, both on zirconium oxidation and on uranium diffusion, is clearly demonstrated. D 2004 Published by Elsevier B.V. Keywords: Uranium diffusion; Irradiation; Zirconium oxidation; Fission products

1. Introduction The use of nuclear power gives rise to a large amount of secondary radioactive nuclear wastes and spent fuel. These wastes have to be rigorously managed, and their safe disposal is a real challenge for the international scientific community. In France, a specific law adopted in December 1991 has proposed an important research program, divided into three areas: (i) separation and transmutation, (ii) deep geological storage and (iii) longterm disposal. This paper is performed in the framework of this last research area and concerns the direct long-term * Corresponding author. Tel.: +33 4 72 43 10 00; fax: +33 4 72 44 80 04. E-mail address: [email protected] (N. Moncoffre). 0257-8972/$ - see front matter D 2004 Published by Elsevier B.V. doi:10.1016/j.surfcoat.2004.08.076

disposal of nuclear spent fuel. The aim of the experiments is to evaluate characteristic parameters of actinide migration in the nuclear fuel cladding tubes made of a zirconium alloy called zircaloy. During nuclear operation, the fuel cladding inner surface is contaminated with fission products and actinides [1]. Due to the temperature effect (around 150 8C), the actinides can migrate through the cladding during the disposal period and their potential radiobiological impact is very important since they are long lived a-emitters. In addition, the diffusion can be enhanced by the defects created by the a irradiation. This diffusion is strongly dependent on the zircaloy oxidation. This is the reason why we have studied the influence of defects on the zirconium oxidation together with the actinide migration.

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2. Uranium diffusion under reactor conditions 2.1. Experimental In order to simulate the evolution of the inner surface of cladding tubes during irradiation, the experiments were performed using the high neutron flux of the Laue Langevin Institute (ILL) of Grenoble. Samples are made of 2-Amthick zirconium substrates on which 300 Ag cm2 uranium oxide layers have been deposited. These UO2 films are enriched up to 98% with 235U. Under the neutrons impacts (5
1014 n cm2 s1), high-energy fission products are emitted and go through the zirconium foil before being detected. Their mass and kinetic energy are analysed with two successive magnetic and electric fields. The separated fragments are identified using a high-resolution ionization chamber [2]. The fission rate reaches 3
1011 particles s1. We have chosen to detect the mass 90 with the most probable charge state ( q=18). Such an experiment [3] has been previously described but without any temperature measurement due to the high radioactive environment in the ILL chamber. In order to solve this problem, a specific infrared pyrometer has been built which allows the measurement of the target temperature at a distance of 15 m, on a 2 cm2 spot and with an accuracy of 10 8C. 2.2. Results At regular time intervals, the energy spectra of the chosen fission fragments are measured, and Fig. 1 represents the evolution of the maximum energy distribution vs. time. As shown in this figure, a first phase corresponding to an energy displacement towards lower values occurs. Then a stabilisation around a mean energy value is observed. We have shown that the first stage corresponds to a full oxidation of the zirconium foil [3]. The zirconium oxide stoichiometry is calculated to be ZrO1.7 from the profile energy shift. The second step corresponds to the diffusion of uranium into the oxide layer. The temperature evolution curve associated with these measurements (Fig. 1) shows

Fig. 1. Evolution of the maximum energy of fission products (A=90, q=18) and of temperature as a function of time.

Fig. 2. Evolution of the energy distribution of fission products (A=90, q=18) during the uranium diffusion phase. Curves have been normalised at the profile maximum, taking into account the burn-up.

that during oxidation, temperature increases progressively from 350 to 470 8C due to the increase of the fission product energy release. It stabilises at 470 8C during the diffusion phase. For further discussion, the beginning of the oxidation process will be taken after 20 h, when temperature starts to stabilise. Fig. 2 presents the evolution of the energy distribution curves after the oxidation phase. A broadening of the curves, which is characteristic of the uranium diffusion process, clearly appears. In order to deduce an apparent diffusion coefficient value from these spectra, the fission product energy distribution curve at the end of irradiation has been simulated. We propose a model where the uranium diffusion follows the Fick’s second law according to BC B2 C ¼D 2 Bt Bx

ð1Þ

where C is the concentration of the diffusing species, and D is the diffusion coefficient. Solutions depend on boundary conditions which are the following: for xN0, C=C(x,t); xV0,

Fig. 3. Fit of the diffusion simulation model (dashed lines) with the experimental fission product energy distribution (A=90). Dots represent experimental data. They are adjusted by Gaussian curves (full lines).

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Fig. 4. Uranium diffusion profile deduced from the presented model.

C=C 0=1/3; x=~, C(~,t)=0; t=0, C(x,0)=0, the analytical solution is   x pffiffiffiffiffi C ð x; t Þ ¼ C0 erfc ð2Þ 2q Dt where C is the atomic uranium concentration, x is the diffusion depth (mg cm2), q is the mass density (mg cm3), D is the diffusion coefficient (cm2 s1), and t is the irradiation time (s). For a given irradiation time t, a D value is supposed from which, according to Eq. (2), one can calculate an uranium diffusion profile. This calculated profile is partitioned in elementary depths dx for which the concentration is supposed to be constant. Each elementary part is associated with a Gaussian distribution. At a given time t, the fission product energy profile is rebuilt by summing the effects of all elementary contributions. The correct D value is obtained by fitting the result of the model to the experimental spectra. An illustration of such a simulation is presented in Fig. 3. The apparent diffusion coefficient deduced from this model is 1015 cm2 s1. Fig. 4 presents the uranium distribution obtained with this diffusion coefficient.

3. Effects of irradiation on zirconium oxidation In order to determine the influence of irradiation defects on the oxidation process, thermal oxidation experiments

Fig. 6. Representation of the thermal oxidation kinetics vs. the square root of time.

were performed in the same pressure condition that the ILL had (5
103 Pa). The oxidation kinetics were followed using the thermogravimetric method. Results are plotted in Fig. 5 at 350, 400, 450 and 480 8C in comparison with the ILL oxidation curve. One observes a classical parabolic thermal oxidation kinetics curve. It is well known that in zirconium, the oxidation process is governed by the oxygen diffusion in the oxide film through the vacancies [4]. The oxidised thickness varies linearly with the square root of time as shown in Fig. 6. Hence, from the curve slopes, oxygen diffusion coefficients in zirconia (ZrO2) have been deduced at each temperature. They are summarised in Table 1 together with literature data [5–9]. The Arrhenius law applied to our data gives an activation energy equal to 1.4 eV atom1 and a D 0 constant of 8.2
105 cm2 s1. Concerning oxidation under irradiation, the process is significantly increased by the irradiation defects as shown in Fig. 5. These defects act as short circuits which accelerate oxidation [4]. However, in order to deduce a rough effect of these defects, we have also supposed here a diffusion

Table 1 Oxygen diffusion coefficients values obtained in zirconia at different temperatures and under irradiation

Fig. 5. Thermal oxidation kinetics at different temperatures compared to that obtained at ILL under irradiation.

Oxygen diffusion coefficient (cm2 s1)

350 8C

400 8C

450 8C

480 8C

P(O2)=105 Under ILL irradiation At atmospheric pressure [5] [6] [7] [8] [9]

3.4
1016

3.0
1015

1.20
1014

3.34
1014 1015

6.7
1015

3.1
1014

1.33
1013

2.37
1013

Comparison with literature data.

1012 3.16
1013 1013 2
1014

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5. Discussion

Fig. 7. Uranium distribution profiles in zirconia as a function of annealing time at 800 8C (U=1016 ions cm2, 800 keV).

phenomenon for oxygen. Considering this approximation, the total oxygen diffusion coefficient at 480 8C obtained at ILL is found to be 1.33
1013 cm2 s1 which corresponds roughly to a factor 4 compared to the thermal value (3.34
1014 cm2 s1). If we write D total=D therm+D irr, then we deduce that D irr=1013 cm2 s1 which points out the major influence of irradiation.

4. Effects of irradiation on uranium diffusion In order to study the influence of defects created by the irradiation on uranium diffusion, experiments of thermal diffusion have been performed. The zirconium oxide was obtained by annealing polycrystalline zirconium foils in air at 450 8C during 5 h. In these conditions, the zirconia layer on to the Zr substrate reaches a 1.5-Am thickness. These samples were then implanted at the Nuclear Physics Institute of Lyon with 238U of 800 keV energy and to the fluence of 1016 ions cm2s1. The uranium ranges calculated with SRIM [10] is 100 nm. Annealings were performed at 700 and 800 8C under primary vacuum (7.5
101 Pa). The study could not be performed in a secondary vacuum because of the zirconia dissolution under such conditions [11]. The uranium concentration profiles were determined using Rutherford backscattering spectrometry (RBS) with 3 MeV a particles. Experimental spectra are displayed in Fig. 7. They lead to the conclusion that in zirconia, even at 800 8C after a 23-h annealing, no broadening of the initial distribution is observed, which means that no diffusion of uranium is detected by RBS, whereas under irradiation, at 480 8C, a uranium diffusion coefficient was determined to be 1015 cm2 s1. Complementary XPS experiments not presented here, performed on thermal annealed implanted samples, shows the appearance of the U 4f5/2 and U 4f7/2, respectively, at 392.7 and 381.8 eV characteristic of U(VI) corresponding to UO3. Without annealing, no metallic uranium signal, U(0), is detected.

We have determined oxygen diffusion coefficients in compact zirconia (before the kinetic transition) between 400 and 480 8C under an oxygen partial pressure of 103 Pa. It has been reported by Nakamura et al. [12] that the diffusion coefficient of oxygen in ZrO2 is proportional to m P O2 , where m is equal to 0.15. Considering data obtained at the atmospheric pressure [5], we find an m value of 0.14, very close to this result. This allows to conclude that the oxygen partial pressure has a low influence on the oxidation rate. If we compare the oxygen thermal diffusion coefficient obtained at 480 8C and under a 103 Pa oxygen partial pressure (D=0.33
1013 cm2 s1) with the result under irradiation in the same temperature and pressure conditions (D=1013 cm2 s1 ), the enhancement due to irradiation corresponds to a factor 3. Considering uranium diffusion, we did not observe any thermal diffusion, even at 800 8C, which means that D is lower than roughly to 1017 cm2 s1, which is the detection limit of RBS. Under irradiation, at 480 8C, a 1015cm2 s1 coefficient is obtained. If we make the hypothesis of a nonthermal diffusion process in this temperature range [13,14], there is at least a factor 100 in the irradiation effect. It could be explained by the fact that for oxygen, the diffusion mechanism through oxygen vacancies is the same, with or without irradiation. For uranium diffusion, the UO3 oxide formation traps the uranium ions in the zirconia lattice. The deposited energy by fission products during irradiation, by increasing the number of defects, prevents this mechanism.

Acknowledgements The authors thank Professor NoJlle Chevarier for her important contribution to this work.

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