MOLTEN SALT TREATMENT AS A DECONTAMINATION METHOD FOR IRRADIATED GRAPHITE

Abstract

In the next decade, all graphite-moderated reactors in the UK are due to close resulting in more than 96,000 tonnes of radioactive graphite waste with variable compositions of long lived (e.g. 14C, 36Cl) and short-lived (e.g. 3H and 60Co) radioisotopes. Due to the presence of the long-lived isotopes, graphite waste is not preferred to be placed in a near surface facility as a long-term strategy and therefore will take up to 30% in volume of the Intermediate Level Waste (ILW) inventory in the UK. Identifying efficient, cost-effective and reliable approach to managing this graphite waste is one of the most critical aspects of any proposed decommissioning strategy for graphite moderated reactors. This work has explored the possibility of decontaminating irradiated graphite using electrolysis in high-temperature molten salt media. Proof of concept studies have been performed using Pile Grade A (PGA) graphite in LiCl KCl eutectic, chosen due to relatively low melting point (625K). The release of corrosion and fission products in molten salt media from the graphite due to electrolysis at 723K was studied. Various process parameters such as up to 80 mA in absolute magnitude of applied current and switching between the reduction and oxidation conditions (electrochemical treatment cycle) up to 10 times were used to optimise radioisotope transfer into the salt phase. By refining the magnitude of current and cycle number, substantial removal of radionuclide contamination (60Co, 133Ba, 137Cs, 154Eu) from the graphite was achieved. Up to 80% reduction of total initial activity for the 60Co isotope was achieved without significant degradation of the graphite material. The magnitude of activity removed from the graphite was sufficient to reclassify the remaining graphite material from ILW to Low Level Waste (LLW). An initial assessment of material’s behaviour and structural changes under molten salt treatment conditions has been conducted using multi-technique characterisation including Brunauer Emmett Teller surface area, scanning electron microscopy, X ray photoelectron spectroscopy and X ray powder diffraction. Limited degradation of graphite material was mainly associated with the enlargement of surface area via partial removal of binder and impregnant phases of the graphite matrix. The stability of lattice parameters pre and post treatment combined with the no significant change in crystalline dimensions indicates no intercalation from molten salt. Molten salt treatment in irradiated graphite has removed up to 15% of 14C and up to 50% 3H, which was found to be increased with applied current/cycling occurrence. This novel molten salt treatment research shows significant decontamination factors of graphite with minimal weight loss or disturbance of graphite material allowing for potential reclassification of the graphite waste stream.

Date of Award	8 Jan 2019
Original language	English
Awarding Institution	The University of Manchester
Supervisor	Clint Sharrad (Supervisor) & Abbie Jones (Supervisor)