Diffusion of radiogenic helium in natural uranium oxides Danièle Roudil a, * , Jessica Bonhoure b , Raphaël Pik c , Michel Cuney b , Christophe Jégou a , F. Gauthier-Lafaye d a CEA Centre de Marcoule BP 17171 30207, Bat 166, BP 17171, 30207 Bagnols sur Cèze cedex, France b Nancy-Université, G2R, CNRS, CREGU, BP 239, 54506, Vanduvre-lès-Nancy, France c CRPG-UPR 2300 CNRS-Centre de Recherches Pétrographiques et Géochimiques, 54506 Vandoeuvre lès Nancy, France d Centre de Géochimie de la Surface, EOST, CNRS/ULP, 1 rue Blessig, 67083 Strasbourg, France article info Article history: Received 14 November 2007 Accepted 6 May 2008 abstract The issue of nuclear waste management – and especially spent fuel disposal – demands further research on the long-term behavior of helium and its impact on physical changes in UO 2 and (U,Pu)O 2 matrices subjected to self-irradiation. Helium produced by radioactive decay of the actinides concentrates in the grains or is trapped at the grain boundaries. Various scenarios can be considered, and can have a sig- nificant effect on the radionuclide source terms that will be accessible to water after the canisters have been breached. Helium production and matrix damage is generally simulated by external irradiation or with actinide-doped materials. A natural uranium oxide sample was studied to acquire data on the behavior of radiogenic helium and its diffusion under self-irradiation in spent fuel. The sample from the Pen Ar Ran deposit in the Vendée region of France dated at 320 ± 9 million of years was selected for its simple geological history, making it a suitable natural analog of spent fuel under repository con- ditions during the initial period in a closed system not subject to mass transfer with the surrounding environment. Helium outgassing measured by mass spectrometry to determine the He diffusion coeffi- cients through the ore shows that: (i) a maximum of 5% (2.1% on average) of the helium produced during the last 320 Ma in this natural analog was conserved, (ii) about 33% of the residual helium is occluded in the matrix and vacancy defects (about 10 5 mol g 1 ) and 67% in bubbles that were analyzed by HRTEM. A similar distribution has been observed in spent fuel and in (U 0.9 ,Pu 0.1 )O 2 . The results obtained for the natural Pen Ar Ran sample can be applied by analogy to spent fuel, especially in terms of the apparent solubility limit and the formation, characteristics and behavior of the helium bubbles. Ó 2008 Elsevier B.V. All rights reserved. 1. Introduction From the perspective of a deep geological repository for spent fuel it is essential to assess the impact of alpha self-irradiation on the long-term stability of the irradiated ceramic. The accumula- tion of irradiation damage and helium by radioactive decay of the actinides could modify the ceramic microstructure after several thousand years, allowing radionuclide release when water comes into contact with the fuel. These changes are likely to be of greater magnitude for high-burnup UO 2 fuel and for (U,Pu)O 2 MOX fuel with high actinide concentrations, especially in plutonium-rich aggregates. These issues were already a subject of investigation in France under the terms of the Bataille Act of 1991 concerning radioactive waste management. The fifteen-year research period specified in that law expired in 2006; a subsequent waste management act in June 2006 temporarily extended the long-term behavior studies of spent fuel in a nuclear waste repository, and highlighted the need for further data on the long-term behavior of helium. With regard to helium behavior, two scenarios directly related to its mobility under alpha self-irradiation at low temperatures can be considered: either the helium remains occluded in the matrix or trapped at the grain boundaries where it could eventu- ally form inter- or intragranular microcracks liable to induce grain decohesion, or the helium is easily released from the cera- mic and can accumulate in the fuel rods. The first scenario is the most challenging in a spent fuel repository because a significant increase in the fuel surface area would have a major effect on the two main source terms: instant release of radionuclides located outside the (U,Pu)O 2 grains and radionuclides located in the (U,Pu)O 2 matrix [1–3]. Many studies of helium behavior and disposition are based on simulated aging methods involving external irradiation by heavy ions or materials doped with short-lived actinides ( 244 Cm or 238 Pu) [4–6]. The basic data on thermal diffusion of helium in UO 2 [7–9] confirm that atomic thermal diffusion is nil over the time and temperature range of a disposal repository or in interim storage. Moreover, some observed phenomena, such as the dis- crepancy between microscopic and macroscopic volume expansion versus the damage level [4], require further investigation of helium behavior and its subsequent evolution. 0022-3115/$ - see front matter Ó 2008 Elsevier B.V. All rights reserved. doi:10.1016/j.jnucmat.2008.05.001 * Corresponding author. Tel.: +33 466 796 172; fax: +33 466 797 708. E-mail address: danielle.roudil@cea.fr (D. Roudil). Journal of Nuclear Materials 378 (2008) 70–78 Contents lists available at ScienceDirect Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat