Solubility of uranium (IV) oxide in alkaline aqueous solutions to 300�C

Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

Most supercritical water-cooled reactor (SCWR) concepts being considered as part of the Generation IV initiative are direct cycle. In the event of a fuel defect, the coolant will contact the fuel pellet, potentially releasing fission products and actinides into the coolant and transporting them to the turbines. At the high pressure (25 MPa) in an SCWR, the coolant does not undergo a phase change as it passes through the critical temperature in the core, and nongaseous species may be transported out of the core and deposited on out-of-core components, leading to increased worker dose. It is therefore important to identify species with a high risk of release and develop models of their transport and deposition behavior. This paper presents the results of preliminary leaching tests in SCW of U-Th simulated fuel pellets prepared from natural U and Th containing representative concentrations of the (inactive) oxides of fission products corresponding to a fuel burnup of 60  GWd/ton. The results show that Sr and Ba are released at relatively high concentrations at 400°C and 500°C.

The Solubility of Unirradiated UO2 In Both Perchlorate And Chloride Test Solutions. Influence of the Ionic Medium

The solubility of a crystalline unirradiated UO2 (s) has been studied under reducing conditions at 25°C in three different ionic media: 0.008 mol dm–3 NaCIO4, 1 and 5 mol dm-3 NaCl+. The species responsible for the solubility in perchlorate medium are U(OH)3 and U(OH)4, with the stability constants log β13 = -0.4(±0.2) and log β14 = -5.7(±0.1), respectively. The solubility in both1! and 5 M chloride media is explained at pH values higher than 6 by the species U(OH)4 with log β14 = -5.5(±0.2) while at lower pH values an abnormal behaviour is observed, with higher uranium concentrations in solution and a proton independent solubility for pH values lower than 4. The XPS observations of the reacted solids do not show the presence of solid surface phases other than uranium dioxide, with an upper oxidation state of UO2.1.

Modelling near- and far-field processes in nuclear waste management

In this chapter we present some of the recent advances concerning process understanding and modelling of radionuclide migration in nuclear waste disposal systems. The present geochemical modelling approaches used to quantify the processes concerning spent fuel dissolution, radionuclide interactions with the canister materials and the bentonite buffer are thoroughly discussed. Finally, some applications to natural analogue studies of spent nuclear fuel disposal, as testing ground for concepts and models developed for waste management systems, are presented.