PNNL

Abstract

Only a handful of high-halide salt waste forms have been demonstrated for vitrification-based immobilization strategies for halide-salt nuclear waste streams (e.g., pyroprocessing wastes, molten salt reactor wastes) and they all have low waste loading potential as well as low chemical durability for high-alkali streams. An alternative approach to direct salt immobilization is salt partitioning prior to waste form fabrication and one option for partitioning is halide removal (dehalogenation). Removing the halogen fraction from through dehalogenation can significantly reduce the waste volume requiring disposal in the primary waste form. When dehalogenation is performed using organic acids, the dehalogenation reagent can decompose during high-temperature vitrification, reducing waste loading limitations in the waste form. In the current work, different organic acids (i.e., oxalic, formic, acetic, oxamic, and citric) were evaluated for dehalogenation efficiency of a simple chloride salt simulant (7.19% LaCl3, 53.77% LiCl, and 39.04% KCl, by mole) and a more complex chloride salt simulant called ERV3 (electrorefiner version 3) at 150°C–300°C and using H+:Cl-molar ratios of 1:1, 2:1, and 3:1. Additionally, a borosilicate glass waste form called TARS (or The Average of Refined Specifications) was formulated for dehalogenated ERV3, with the waste form properties characterized.