Uranium (IV) Dioxide and Simfuel as Chemical Analogues of Nuclear Spent Fuel Matrix Dissolution. A Comparison of Dissolu
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URANIUM (IV) DIOXIDE AND SIMFUEL AS CHEMICAL ANALOGUES OF NUCLEAR SPENT FUEL MATRIX DISSOLUTION. A COMPARISON OF DISSOLUTION RESULTS INA STANDARD NaCIINaHCO 3 SOLUTION
JORDI BRUNO*, I. CASAS**, E. CERA*, J. DE PABLO**, J. GIMIENEZ** AND M.E. TORRERO**;
Tecnologia Ambiental, 08290 Barcelona, Spain. **Dept. Chem. Eng., Universitat Politecnica de Catalunya, 08028 Barcelona, Spain. * MBT
ABSTRACT We have carried out an experimental comparison study of the dissolution rates of unirradiated U02 and SIMFUEL pellets and particles (100-300 plm) in a standard NaCI/NaHCO 3 solution, under oxidizing conditions. We have performed the experiments using batch and flow methodologies. Both methodologies gave similar results, indicating that the overall oxidation/dissolution process is the same in both cases. The results from the experiments indicate that under these conditions the dissolution process is both oxygen and bicarbonate promoted. The dissolution rates we obtained are: R=2.4 ± 0.8 mg U/51 d for U02 and R= 0.17 ± 0.05 mg U/m2 d for SIMFUEL. The results of the experiments indicate that the dissolution rate under oxic conditions is clearly dependent on the number of U(VI) surface sites which for spent nuclear fuel is a function of the extent of radiolytic oxidation. INTRODUCTION The dissolution of the spent fuel matrix constitutes the largest source of radionuclides under repository conditions [1]. Consequently the stability of the matrix and its rate of dissolution are critical parameters to assess the performance of a spent nuclear fuel repository. Ideally, it would be possible to devise an extensive research program on the dissolution of U02 spent fuel to derive all the critical thermodynamic and kinetic parameters required. Unfortunately, this is neither technically nor economically feasible. In this context, the use of chemical analogues to the waste matrix, as U02 and SIMFUEL can be useful. The validity of the use of these chemical analogues [2,3] has been proven and their limitations discussed [2]. The usefulness of the dissolution experiments is also dependent on the experimental methodology used to perform them. In this context it is important to ascertain to which extent the dissolution test methodology (batch or continuous flow) affects the information obtained to be used in the performance assessment of spent nuclear fuel. As an example, higher dissolution rates have been obtained under oxidizing conditions, in experiments performed by using continuous flow [4]. The effect of the carbonate on the dissolution rates is another important parameter. Dissolution rates are higher on carbonate solution, as it has been pointed out by several authors [5]. However, we do not know if this is the result of a kinetic enhancement or due to an increase in the solubility limit. A comparison between batch and flow experiments under the same conditions may give us some clues about these matters. Therefore, we have studied experimentally the dissolution rates of both U02 and SIMFUEL under oxidizing conditions using a standard so
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