Behavior of Np, Pu, Am, Tc Upon Glass Corrosion in a Concentrated Mg(Ca)Cl 2 Solution
- PDF / 512,445 Bytes
- 8 Pages / 414.72 x 648 pts Page_size
- 75 Downloads / 169 Views
BEHAVIOR OF Np, Pu, Am, Tc UPON GLASS CORROSION IN A CONCENTRATED Mg(Ca)C12 SOLUTION Bernd Grambow, Andreas Loida, Lothar Kahl, Werner Lutze*, Kernforschungszentrum Karlsruhe (KfK), Germany; *The University of New Mexico, Albuquerque NM, USA. ABSTRACT The objective of this investigation is to describe the extent to which Np, Pu, Am and Tc are mobilized from vitrified high-level radioactive waste into the near field of an HLW repository in a salt formation, when a hot and concentrated salt solution comes into contact with the glass. Waste form corrosion studies are conducted with a salt solution representing the composition of a fluid phase encountered in drill holes in the Gorleben salt dome. Test temperatures are determined by the designed maximum surface temperature of 200'C for the vitrified waste in the Gorleben salt. The following results were obtained: 1. pH changes of the radio-active leachate are the same as in inactive leachates. 2. The time and temperature dependence of the reaction for the radioactive glass are in excellent agreement with that of the inactive glass. 3. Np, Pu, Am, and Tc have not been reimmobilized in secondary minerals. Hence, mobilization of these radionuclides is governed by the kinetics of glass dissolution. Pu oxidation states were analyzed and related to Pu concentrations. INTRODUCTION Characterization of the chemical durability of waste forms containing simulated HLW must be complemented by studies with radioactive samples in order to investigate radiation effects and the behavior of actinides The objective of this investigation is to describe the extent to which Np, Pu, Am and Tc are mobilized from vitrified high-level radioactive waste into the near field of an HLW repository in a salt formation, when a hot and concentrated salt solution comes into contact with the glass. Waste form corrosion studies are conducted with a salt solution representing the composition of a fluid phase encountered in drill holes in the Gorleben salt dome. The maximum test temperature is determined by the designed maximum surface temperature of 200'C for vitrified waste in the salt.
EXPERIMENTAL Glass samples containing reprocessing waste were melted by CEA's 'Centre de la Vall6e du Rh6ne', France, and shipped to KfK. Radionuclide concentrations are given in table 1. The glass composition is similar to that of the COGEMA glass R7T7 produced in the vitrification plants 'R7 and T7 at La Hague, France. Activity concentrations, except for Pu, are lower in the CEA glass than in the COGEMA glass, because a different waste was used by CEA. The composition of the inactive CEA glass R7T7 can be found in [1]. The CEA glass was powdered and an average grain size of 86 prm was sieved out and corroded in a Halite saturated concentrated Mg(Ca)C1 2- salt solution for periods of time up to 1 450 days at 1100, 1500, and 190TC (surface area to solution volume ratio S/V = 9370 m- ). The composition of the salt solution is given in [2]. Fully tantalum-lined autoclaves with graphite seals were used. The concentrations of
Data Loading...
