Radioiodine released during the nuclear‑fuel cycle constitutes a persistent radiological hazard. In this study, the IO3- uptake mechanism of CoAl LDH was resolved by combining pH‑controlled sorption experiments, synchrotron XAFS, and DFT-based AIMD simulations. At pH close to 6, approximately 90 % of IO3- was removed, and the equilibrium distribution coefficient reached about 1.7 × 104 mL g-1. EXAFS analysis indicated an average iodine-oxygen bond length of 1.81 Å and a coordination number near 3, with the fit R-factor equal to 0.002. The simulations faithfully reproduced the experimental spectrum and revealed transient proton hopping events that generated metastable I-O-H species inside the interlayer, thereby confirming nitrate-to-iodate exchange as the controlling capture pathway. Atomic density profiles and radial distribution functions further showed that IO3- adopt an end‑on orientation perpendicular to the hydroxide sheets, while water molecules mediate proton migration without disturbing the host lattice. The integrated experimental-computational evidence demonstrates that CoAl LDH can rapidly and selectively sequester IO3- under near‑neutral conditions, offering atomic scale guidance for the rational engineering of layered sorbents for advanced radioactive‑waste treatment.

Radioactive iodine isotopes (129I and 131I) from spent nuclear fuel pose significant environmental risks due to high radioactivity and mobility in aqueous systems. This study embedded NiAl Layered Double Hydroxide (LDH) within sodium alginate and poly vinyl alcohol matrices by crosslinking with CaCl2 to fabricate bead-type sorbents for I- removal. XRD and FT-IR analyses confirmed that the crystallinity of NiAl LDH was retained within the beads, indicating structural stability. However, the sorption capacity of NiAl LDH beads (0.2151-0.2489 mmol & centerdot;g-1) was lower than that of pristine NiAl LDH powder (0.6750 mmol & centerdot;g-1), primarily due to partial anion-exchange of interlayer NO3- by Cl- during bead formation, as Cl- has a higher affinity than NO3-. Despite this, effective I- sorption occurred. Zeta potential measurements revealed an increase in surface potential after I- sorption, which contradicted the typical behavior of electrostatic attraction. This suggests that structural rearrangement of the bead, driven by Na+-Ca2+ exchange under NaI used for I- solution, may have led to increased LDH surface exposure. This exposure enabled I- sorption via anion-exchange, allowing partial substitution of interlayer anions. These findings can offer insights for the design of bead-type sorbents optimized for radioactive iodine removal.

Bentonite is regarded as an adequate buffer material in deep geological repositories and its swelling properties serve to prevent the penetration of groundwater into the repository and to minimize the release of radionuclides. However, bentonite is rarely effective in removing anionic radionuclides due to its permanent negative surface charge. The aim of this study was to enhance the anion removal ability of bentonite by incorporating layered double hydroxides (LDH) with a high anion exchange capacity. The functionalization of CuAlBi LDH on bentonite (CuAlBi LDH@Ben) revealed an effective approach for removing both cesium and iodine from aqueous solutions. The peak shift of the Si–O stretching band to higher frequencies, the vertically oriented platelet morphology, and the increase in specific surface area provide confirmation that LDH platelets grow on the surface of montmorillonite. The CuAlBi LDH@Ben demonstrates enhanced anion retention performance in bentonite without impacting its retention behavior toward cations, as evidenced by Kd values of 1943.1 mL/g for Cs+, 442.4 mL/g for I−, and 650.7 mL/g for IO3−, respectively.

Radioactive nickels are significant environmental concerns because of their long half-lives (59Ni = 76,000 years and 63Ni = 100.1 years) and high proportions in nuclear wastes in the repositories. Therefore, the investigation of sorption and mobility of Ni at the disposal facility sites is important for the prediction and evaluation of radiological risks to the public. Herein, the batch sorption and column experiments were performed under various geochemical conditions using site-specific rock and groundwater samples collected at the Wolsong low and intermediate level waste (LILW) disposal facility site in South Korea. The batch sorption results revealed that the sorption of Ni was strongly affected by the pH. Approximately 20% Ni adsorbed at pH 5, which enhanced over 90% at pH > 8. The batch sorption results were used to develop the surface complexation model (SCM) to predict the Ni sorption and its mobility behavior. The non-electrostatic generalized composite (GC) SCM approach simulated the Ni sorption data well. We believe, our study can interpret and extrapolate the sorption and transport behavior of Ni in the underground repository conditions.

Technetium-99 ( 99 Tc), a radionuclide generated from nuclear industry is a great environmental concern because of its long half-life (2.13 × 10 5 years) and high mobility in environment. Therefore, apposite management of 99 Tc is imperative to control its hazardous radiological impact on humans and other livings. So far, the major strategy implementation has been the solidification and immobilization of 99 Tc radioactive waste in various matrices as waste forms and disposal in deep geological repository. However, by passing the time, 99 Tc may leach out/solubilize from the waste forms under different geochemical/environmental conditions. In this minireview, we discuss some key contributions towards the solubility of 99 Tc and rhenium (Re; a well-known surrogate of 99 Tc) from different waste forms. Specifically, we review the solubility of 99 Tc from glass, cement, ceramic, and geopolymer waste forms. The final section (conclusion) presents a short summary and future challenges need be addressed to impede the solubility of 99 Tc from the designed waste forms. We believe this minireview will be beneficial to provide a significant insight on the solubility of 99 Tc from aforementioned waste forms and in the design of robust matrices to minimize/prevent 99 Tc migration in various environments.

PuO2(cr) dissolution in natural water was investigated at 25°C and 60°C under atmospheric conditions. The concentration of Pu in solutions [Pu], was monitored for 1 year of reaction time. PuO2(cr) dissolution in natural water reached a steady state within 2 months at 25°C. The [Pu] in groundwater and seawater at pH 8 were in the range of [Pu] = 0.9–34 and 3.4–27 nM, respectively. The [Pu] in concrete porewater (rainwater equilibrated with concrete) at pH 8.1–10.9 was in the range of 0.1–3.2 nM. The [Pu] and pH values of groundwater were similar to those of seawater samples having a high ionic strength. The measured [Pu] at equilibrium in all samples was higher than the calculated solubility curves for PuO2(am, hyd). Experimental evidence is insufficient to confirm the oxidation state of Pu in solution and solid phases. However, the results of geochemical modeling indicate that PuO2(am, hyd) and aqueous Pu(IV) species are dominant in natural water samples of this work. The dissolution behavior of PuO2(cr) in natural waters is comparable to the oxidative dissolution of PuO2(am, hyd) in the presence of PuO2(coll, hyd). The dissolution of PuO2 in groundwater decreased at higher temperatures, whereas the influence of temperature in seawater and porewater was not significant under these experimental conditions.

Ion exchange resins (IERs) are widely used to remove radioactive contaminants from various commercial nuclear power plant systems. After the completion of their useful life cycle, IERs are removed from the nuclear facility and are known as spent resins. In the past few decades, the development of Fenton/Fenton-like treatments for the management of spent resins has received considerable attention because of their potential to completely break down the IER structure into harmless compounds (e.g., carbon dioxide, water, and inorganic salts). In addition, after Fenton-like treatments, the resulting solutions containing radionuclides can be easily immobilized to stable waste. In this review, we critically discuss the key developments in Fenton/Fenton-like dissolution, degradation, and mineralization of spent resins. We describe the important reaction parameters (initial pH, resin dosage, catalyst type and dosage, hydrogen peroxide dosage, flow rates of the catalyst and oxidant, reaction temperature, treatment time, and other specific parameters) for various Fenton/Fenton-like treatments of spent resins. Moreover, this review focuses heavily on the major reaction intermediates generated in Fenton-like treatments. In the final section of this review (conclusions and perspectives), we discuss the major challenges and suggest future research directions need to be addressed to improve the efficiency of Fenton-like treatments of spent resins.

Bentonite is the most probable candidate to be used as a buffer in a deep geological repository with high swelling properties, hydraulic conductivity, thermal conductivity, and radionuclide sorption ability. Among them, the radionuclide sorption ability prevents or delays the transport of radionuclides into the nearby environment when an accident occurs and the radionuclide leaks from the canister, so it needs to be strengthened in terms of long-term disposal safety. Here, we proposed a surface modification method in which some inorganic additives were added to form NaP zeolite on the surface of the bentonite yielded at Yeonil, South Korea. We confirmed that the NaP zeolite was well-formed on the bentonite surface, which also increased the sorption efficiency of Cs and Sr from groundwater conditions. Both NaP and NaX zeolite can be produced and we have demonstrated that the generation mechanism of NaX and NaP is due to the number of homogeneous/heterogeneous nucleation sites and the number of nutrients supplied from an aluminosilicate gel during the surface modification process. This study showed the potential of surface modification on bentonite to enhance the safety of deep geological radioactive waste repository by improving the radionuclide sorption ability of bentonite.

The treatment of radioactive iodine released from nuclear power plants and radiological waste disposal sites is of great concern due to its high mobility and toxicity. In particular, iodide (I-) and iodate (IO3-) are the major iodine species of concern under various pHs and groundwater conditions. Herein, CoAl and NiAl layered double hydroxides (LDHs) were synthesized and investigated to identify the iodine removal mechanism and efficiency. Both CoAl and NiAl LDHs exhibited rapid iodine removal processes within 20 min, following the pseudo-secondorder model via ion-exchange with parent NO3- anion in the LDHs. The CoAl LDH's maximum sorption capacities for I- and IO3- were about 1.67 and 2.16 mmol g(-1), respectively, whereas for the NiAl LDH, these were about 2.10 and 2.26 mmol g(-1), and they followed the Langmuir isotherm model. Interestingly, both the CoAl and NiAl LDHs showed a preferential ion-exchange affinity for IO3- over I-, which was attributed to the structural similarity of the IO3- and NO3- as well as new formation of secondary Co(IO3)(2)center dot 2H(2)O or Ni(IO3)(2)center dot 2H(2)O phases. In addition, a desorption study indicated that the selectivity order was SO42- >= IO3- >= OH- > HCO3- > Cl- > NO3- >= I- and demonstrated the higher retention of the IO3- than I- anion. This study provides insights into promising iodine sorbents and the different removal mechanisms of I- and IO3- using CoAl and NiAl LDHs.

Radioactive cesium and iodine have become an emerging issue because these radionuclides have high radio toxicity and solubility, and they can easily transport to the surrounding environment. Prussian blue, known as metal hexacyanoferrate compound, has a cubic face-centered structure and similar ionic size to Cs+, known to be used for Cs+ removal. In addition, layered double hydroxides (LDHs) have anion exchange properties and become the promising sorbents of iodine. In this study, based on the advantages of both Prussian blue (PB) and layered double hydroxide (LDH), Prussian blue functionalized layered double hydroxide (PB-LDH) was synthesized to concurrently remove both Cs+ and IO3(- )from radioactive solutions. During the synthesis of PB-LDH, Fe(CN)(6)(4-)does not intercalate into the LDH structure, but aggregates on the LDH surface according to electrostatic force and ion-specific effects and forms the PB on the surface of LDH. The synthesized PB-LDH simultaneously removed 32 mg/g of Cs+ and 91 mg/g of IO3- from the mixed waste solutions. The main removal mechanism of Cs+ was redox coupled alkali metal ion transport to form Cs4Co4Fe3II(CN)(18) species at the surface of PB-LDH. However, IO3- was mainly removed by ion exchange with intercalated Cl- in addition to the redox effect of Co ions.

We report the size effect of ion exchange resins (IERs) on Cs and Co distribution in polymer waste forms. Ball mill ground IERs (BG) waste form resulted in relatively better homogeneous waste distribution and displayed superior Cs and Co leachability indexes compared with the same polymer waste form prepared with non-ground IERs (NG).

Radioactive iodine-129 (I-129) is a contaminant of concern at radiological waste disposal and nuclear power production sites where it can incidentally be released into the environment and migrate to groundwater. Development of remediation options for 129 1 is hindered by its complex and mobile redox speciation. In this work, we systematically evaluated nanostructured layered double hydroxide (LDH) materials comprised of MgFe or CoCr with variable M2+:M3+ molar ratios for their affinity for the environmentally-relevant anions of iodine including iodide (I-) and iodate (IO3-). Uncalcined MgFe material demonstrated overall poor selectivity, however, affinity for I- significantly increased with increasing Mg:Fe molar ratio and upon calcination due to elimination of the parent anions followed by reconstruction of the original LDH structure, assuming physisorption. Specific surface area of calcined MgFe before and after I- sorption showed 44.3 and 13.2 m 2 g(-1), respectively, indicating significant decrease by reconstruction effect. In addition, high sorption capacity of calcined MgFe for I- at about 2.5 mmol g(-1) suggests its applicability for immobilization of 129 1. On the other hand, uncalcined CoCr material exhibited high selectivity, affinity, and capacity (about 2 mmol g(-1)) for IO3- indicating chemisorption and its utility for treatments of I-129 in groundwater. There was a minor specific surface area variation of uncalcined CoCr before and after IO3- sorption by ion exchange (53.2 and 64.0 m(2)g(-1), respectively). Overall this study provides mechanistic insights into understanding and prediction of behavior of the LDH materials as efficient sorbent for design of remedial option for I-129.