Thermal diffusion of Helium and volatil fission products in UO2 and zirconolite nuclear ceramics. Danièle Roudil1 ; Xavier Deschanels1 ; Patrick Trocellier2 ...
Thermal diffusion of Helium and volatil fission products in UO2 and zirconolite nuclear ceramics Danièle Roudil1 ; Xavier Deschanels1 ; Patrick Trocellier2 ; François Jomard3; Annick Boutry3; Christophe Jégou1; Sylvain Peuget1; Dominique Gosset4, Pierre Nivet1 1 CEA VALRHO DEN/DTCD/SECM/LMPA, BP 17171 30207 Bagnols sur Cèze cedex, France 2 CEA SACLAY DEN/DMN/SRMP 91191Gif-sur-Yvette cedex, France 3 CNRS Meudon-Bellevue, Place A.Briand, 92195 Meudon , France 4 CEA SACLAY DEN/DMN/SRMA 91191Gif-sur-Yvette cedex, France ABSTRACT The behaviour and diffusion mechanisms of helium in nuclear ceramics, such as uranium dioxide spent fuel matrix and zirconolite for the specific conditioning of minor actinides, significantly impact the possible evolution of those matrices in interim storage or disposal conditions. In the framework of spent fuel storage studies, the additional diffusion of gas and fission products in uranium dioxide matrix is also an essential aspect of the R&D. Specific experimental studies have been conducted, devoted to thermal diffusion under 1000 C. Data processing methods, lead to helium diffusion coefficient and associated activation energy of 1.05 eV in the zirconolite and 2 eV in UO2. Comparatively with the uranium dioxide matrix, the helium diffusion coefficient in zirconolite is 1 to 100 million times higher; this parameter will have to be taken into account to dimension the waste form. Diffusion coefficients measurements between 800 C and 1000 C, investigated by SIMS in the CNRS Meudon, showed a very slow diffusion of volatile fission products Xe, I, Te and Cs, with coefficients two or three order of magnitude lower than for helium. INTRODUCTION The issues of high-level radioactive waste management are extensively investigated in France within the framework of the December 1991 law for the management of spent fuel in the backend of the cycle. As part of the third research area devoted to long term interim storage and geological disposal of spent nuclear fuel, one of the major operational R&D questions concerns the monitoring of spent fuel packages in storage and the applicability of potential reconditioning processes. One of the consequences arising from the presence of actinides in these materials is the formation of a large quantity of helium produced by α disintegration. This amount of gas with very low solubility in many ceramics is to be added, for UO2, to insoluble fission gas (Xe, Kr) created under radiation in reactor.This phenomenon leads to physical modifications in the matrix accompanied by a significant variation in the source term that must be taken into account to evaluate the radionuclide balance liable to be released if the containment is breached. Our study, following a classic procedure, has consisted of three steps ([1], [3])): • implantation of ions in UO2 and zirconolite samples, • annealing of samples at various temperatures to induce helium diffusion, • measurement of the helium profile in the material by the resonant 3He(d,p)4He nuclear reaction or by SIMS for F.P.
EXPERIMENTAL Sample characteristics The characteristics of the sample used are summarize in table I. Zirconolite pellets of composition Ca0.8RE0.17Zr0.94Ti1.86Al0.14O7 (RE = Rare Earth) were provided to us by the Australian Nuclear Science and Technology Organisation. Depleted polycrystalline UO2 pellets were fabricated by sintering (see on figure 1 SEM observations). Ions were implanted in the materials at room temperature at the IPN of Villeurbanne .Table II give the characteristics of the implantation. Depth and maximum concentration of the implantation peak are calculated by SRIM 2000 ([2]). Helium concentrations of this magnitude would be obtained after several hundred years of storage in UOX or MOX fuel and after several thousand years in zirconolite. Those F.P. concentrations are representative of the amount of xenon and caesium in spent fuel at high burn up. Determination of implanted ions profiles and digital processing of spectral data The helium-3 profile was measured by the 3He(d,p)4He nuclear reaction using the nuclear microprobe in the CEA’s Pierre Süe Laboratory at Saclay. This reaction has a maximum cross section for a deuteron energy of 450 keV. During the analysis, the incident deuteron beam energy was progressively reduced from 1500 to 900 keV in variable steps to probe the entire helium-3 profile. The protons produced by the 3He(d,p)4He reaction had an energy of about Table I : Sample pellets characteristics UO2 8.19 10.96 94 ≈10
Mean diameter (mm) Theoretical density (g/cm3) Mean geometric density (% theorical) Grain size (µm)
Zirconolite 8 4.67 93
