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International audience; This paper describes the methodology we have developed to define a sampling strategy adapted to operational constraints in order to characterize the dihydrogen flow rate of 2714 nuclear waste drums produced by radiolysis reaction of organic mixed with α-emitters. The objective was to perform few but relevant measurements. Thus, a sample of only 38 drums has been selected to be measured. Statistical analysis of drum measurement data of dihydrogen rate provided an estimation of the mean and the upper bound of the physical quantity of interest which gave a good convergence with global measurements from the ventilation system of the facility. Thereafter, performing a factorial data analysis has demonstrated the representativeness of the measurement data set and the sampling strategy assumption validity. Moreover, it provided information that has been used for a regression analysis to develop a linear prediction model of dihydrogen flow rate production for the waste drum characterization.

This paper documents the results of a study on the relationship of the pore structure and the transport properties of potassium- and sodium-based geopolymers. The pore structure was investigated by water porosity, mercury intrusion porosimetry and thermoporometry. The geopolymers transport properties were determined using a set of experiments, including natural diffusion of tritiated water and two electrical methods (direct current and Electrochemical Impedance Spectrometry). The geopolymers microstructure corresponds to materials with a pore structure made of poor access dimensions around 7 nm (potassium-based geopolymers), and 20 nm (sodium-based geopolymers) connected to larger pores not detected by MIP but revealed by thermoporometry. As a result, the transport properties appear to be higher than expected.

This work documents the modeling of ionic transfer through porous materials under an external electrical field. A pore model network that accounts for the electrical double layer impact was developed and scaled up at the porous material level. The model is adapted to two different materials, both belonging to the class of mesoporous materials: a monomodal porous glass, and a cement paste with mineral additions. The comparison between numerical results and experiments on migration and tritium natural diffusion questions the concept of formation factor applied to such mesoporous materials.

International audience; The decommissioning of French gas-cooled nuclear reactors, all stopped since 1994, will generate 23,000 tons of nuclear waste containing 14C among other radionuclides. This work aims at presenting and testing a selective extraction treatment of this radionuclide from graphite waste by carbon dioxide gasification.In previous works, we highlighted that graphite turns into a heterogeneously degraded carbon with neutron irradiation and 14C could preferentially be concentrated in the most degraded zones. Moreover, these last react selectively with CO2 that would allow the 14C extraction from graphite waste.The results on graphite waste are promising with the quarter of 14C inventory extracted for a relative weight loss of few percent. Up to 68% of 14C inventory was extracted, but with a more important mass consumption. Thus, the potential of this selective 14C extraction would allow imagining alternative scenarios for graphite waste management.

WOS:000359166000013; International audience; The diffusion coefficient of tritium in its liquid form through cementitious materials is an indicator of durability in the domain of nuclear waste disposal. Because the cement-based materials designed for the waste storage have very high confinement properties, diffusion tests tend to last several months if not years in some cases. Therefore, basing the determination of the tritium diffusion coefficient on the experimental data measured during non steady state is tempting. The aim of this work is to demonstrate that the determination during the transient regime of the effective diffusion coefficient and the material porosity leads to non-negligible underestimation of these quantities. Moreover, it appears that, when using a numerical model for the determination, their validity can only be judged on the basis of more than one criterion, whether those criteria are obtained from the experimental data or from a numerical model. (C) 2015 Elsevier Ltd. All rights reserved.

WOS:000320036400011; International audience; The use of cement-based materials for radioactive waste confinement and storage must rest on precise measurements of their physical and chemical properties. An important property is the diffusivity of tritium in its liquid form (tritiated water) through a sample considered as representative. Here, we report quantitatively the effect of radioactive decay compared to the effective diffusion coefficient on tritium diffusion. The numerical model was validated by comparing it to experimental data. We found that, in the worst case scenario, the calculated effective diffusion coefficient of tritiated water based on the classical analytical solution to Fick's law is underestimated by more than 20 %, compared with the results provided by the numerical model, which accounts for the radioactive decay within the material.

In water-saturated media the main mode for contaminant transport is liquid transfer, and the confinement capacity of the materials is notably characterized by the effective diffusion coefficient (EDC) of tritiated water. The major problem lies in the duration of experiments, that can exceed several years, so two methods have been explored for the development of accelerated ones. The first consists in a variation of the through-diffusion technique: a given tracer concentration, C 0 , is applied on one face of the sample that has been previously impregnated with a C 0 /2 concentration. The duration of the essay can then be divided by 3. The second method involves accelerated migration under the influence of an electric field and direct measurement of the current density. The results are in the same range as the classical through-diffusion experiments, and obtained in about one month. In non water-saturated media, the diffusing fluid of interest is generally the gaseous phase. Two applications at various steps of the nuclear fuel cycle are presented: characterization of the migration of hydrogen in the host rock formation of a geological waste storage, and of the diffusion of tritium gas in the concrete containment structure of decommissioned UNGG nuclear power plants. In both cases the media are close to saturation and the pore water content must be precisely controlled. This paper presents a method which allows to determine the intrinsic permeability and gas diffusion coefficients of the materials.