Development of sorbent for radioactive iodine removal and safety assessments of iodine and beryllium in disposal environment

Abstract

Bismuth functionalized graphene oxide (Bi-GO) was successfully synthesized and tested for iodine removal. A high removal efficiency for both iodide and iodate from radioactive wastewater was obtained. Batch experiments for kinetic and selectivity tests were carried out, respectively. Additional SEM, XRD, FT-IR, and XPS analyses were conducted to characterize the sorbent and identify the presence of bismuth on the GO surface. It was confirmed that the bismuth on the GO surface reacts with the iodine species by surface complexation (or precipitation). The dominant surface species of iodide and iodate were confirmed by BiOI and Bi(IO3)3, respectively. After the selectivity test using a KCl background solution with varying concentrations (1, 10, and 100 mg·L−1), the Bi-GO still showed higher removal efficiency (≥ 95%) for both iodide and iodate than that of the commercial silver exchanged zeolite (≥95% for iodide and ≤25% for iodate). This study provides a potential application of Bi on graphene-based materials for selective removal of both iodide and iodate species from radioactive wastewater. Beryllium is a chemotoxic element used in nuclear reactors as the reflectors or moderators due to its low thermal neutron absorption cross-section and its specific chemical/structural properties. Consequently, beryllium is also found in specific radioactive wastes in repositories. In this regard, this study focuses on the sorption of beryllium on fresh Portland cement (degradation phase I, pH ≈ 13) under an Ar-glove box at T = (22 ± 2) °C. Ordinary Portland cement (CEM I 42,5 N BV/SR/LA type) was used as sorbent material in all the experiments. Batch sorption experiment were conducted with 10-6M ≤ [Be(II)]0 ≤ 10-2.5 M and 2 g·L-1 ≤ [S/L] ≤ 50 g·L-1, where the solubility limit of β-Be(OH)2(cr) was considered in the definition of [Be(II)]0. A relatively strong uptake of Be(II) by ordinary Portland cement in the first degradation phase. The GoldSim simulations were employed for the post disposal safety analysis of the shallow depth type repository in Korea. This program is used for both deterministic and probabilistic total system performance assessment; it can evaluate the nuclide release from the repository. Three scenarios, inner site of the repository, near field, and far-field were developed for the safety simulation and the chemical and physical factors for input data were calculated and obtained experimentally, respectively. The calculated radiotoxicity flux and activity from the repository inner site, near-field, and far-field attained the maximum value of 1.15e-7, 8.05e-13, and 2.44e-5 Sv/m2/yr for iodine and 5.72e-7, 4.011e-12, and 8.73e-5 Sv/m2/yr for beryllium, respectively. The results showed that radionuclides released from shallow depth type repository were retarded and influenced by chemical adsorption of barrier systems in the environment.