• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1218-1223
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• 2021

A hydrophobic ionic liquid including an amino moiety ([DiOcAPmim][NTf₂]) was synthesized. Its extraction behaviors towards Pd(II), Ru(III), Rh(III) were investigated in nitric acid aqueous solution as a function of contact time, effect of concentration of nitric acid, effect of temperature, and effect of co-existing metal ions. The extraction kinetics of Pd(II) was fairly fast and extraction equilibrium can be attained within only 5 min under the [HNO₃] = 2.05 M. When [HNO₃]< 1 M, the extraction percentage of Pd(II), Ru(III), Rh(III) were all above 80%. When [HNO₃] reached 2 M, all of the extraction percentage decreased and in an order of Pd(II)>Ru(III)>Rh(III). When [HNO₃]> 2 M, the extraction performance gradually recovered. The effect of temperature can slightly affect the extraction performance of Pd(II). Furthermore, in simulated high-level liquid waste, [DiOcAPmim][NTf₂] showed a better preference towards Pd(II) under the interference of various other co-existing metal ions.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3641-3649
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• 2022

A simple solvothermal reaction was used to prepare a 3-aminopropyl-functionalized silica-gel-based adsorbent for adsorbing Pd(II) from the nitric acid solution. Scanning electron microscopy, fourier transform infrared spectroscopy, and thermogravimetry analysis were performed on the as-synthesized adsorbent to demonstrate the successful introduction of Schiff base groups. Batch experiments were used to investigate the effects of contact time, nitric acid concentration, solution temperature, and adsorption capacity. It is worth noting that the prepared adsorbent exhibited a higher affinity toward Pd(II) with the uptake approximately 100% even in a 2 M HNO₃ solution. At an equilibrium time of 5 h, the maximum adsorption capacity of Pd(II) was estimated to be 0.452 mmol/g. The adsorbed Pd(II) could be completely eluted by dissolving 0.2 M thiourea solution in 0.1 M HNO₃. Using a combination of particle-induced X-ray emission analysis and an X-ray photoelectron spectrometer, the adsorbed Pd was found to be uniformly distributed on the surface of the prepared adsorbent and the existing species were Pd(II) and zero-valent Pd(0). Due to the desirable performances, facile preparation method, and abundant raw material source, the prepared adsorbent demonstrated a high application potential in the recovery of Pd(II) from simulated high-level liquid waste treatment.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.812-818
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• 2024

The volume and toxicity of radioactive waste can be decreased by separating the components of high-level liquid waste according to their properties. An impregnated silica-based adsorbent was prepared in this study by combining N,N,N',N',N",N"-hexa-n-octylnitrilotriacetamide (HONTA) extractant, N',N'-di-n-hexyl-thiodiglycolamide (Crea) extractant, and macroporous silica polymer composite particles (SiO₂-P). The performance of platinum-group metals adsorption and separation on prepared (HONTA + Crea)/SiO₂-P adsorbent was then assessed together with that of co-existing metal ions by batch-adsorption and chromatographic separation studies. From the batch-adsorption experiment results, (HONTA + Crea)/SiO₂-P adsorbent showed high adsorption performance of Pd(II) owing to an affinity between Pd(II) and Crea extractant based on the Hard and Soft Acids and Bases theory. Additionally, significant adsorption performance was observed toward Zr(IV) and Mo(VI). Compared with studies using the Crea extractant, the high adsorption performance of Zr(IV) and Mo(VI) is attributed to the HONTA extractant. As revealed from the chromatographic experiment results, most of Pd(II) was recovered from the feed solution using 0.2 M thiourea in 0.1 M HNO₃. Additionally, the possibility of recovery of Zr(IV), Mo(VI), and Re(VII) was observed using the (HONTA + Crea)/SiO₂-P adsorbent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.245-256
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• 2004

The extraction of three kinds of amido podands, N,N,N'N'-tetrabutyl-3-oxa-pentanedi- amide (TBDGA), N,N,N'N'-tetra-isobutyl-3-oxa-pentanediamide(TiBDGA) and N,N,N'N'-tetra- butyl-3,6-dioxa-oct-anediam- ide(TBDOODA) on U(VI),Pu(IV), Am(III), Eu(III) and other metal ions is studied in nitric acid solutions. 40%octanol-kerosene is chosen as diluents to eliminate third phase and emulsion. TBDGA and TiBDGA show extraction selectivity to An(III) and Ln(III) much higher than to U(VI) and Pu(IV). Fe, Ru and Mo is poorly extracted by the three kinds of amid podands in 2~3mol/L $HNO_3$ solutions. Aiming to eliminate interface crude when using simulated HLLW solution in the system of 0.2mol/L TBDGA/Octanol＋kerosene, acetohydroxyamic acid was adapted. Distribution ratio of zirconium was decreased when adding acetohydroxyamic acid in aqueous solution, and interface crude disappeared as mixing extractant with HLLW. The counter-current extraction test is carried out in a set of miniature mixer-settler, with 0.2mol/L TBDGA/ 40% octanol-kerosene as extractant to separate U(VI), Pu(IV), Am(III) and Eu(III) from simulated high level liquid waste(HLLW) solution. In battery A, lanthanides and actinides are coextracted into organic phase with the recovery of 99.98% for U(Ⅵ), >99.99% for Pu(IV), and >99.99% for Am(III) and Eu(III) respectively. In battery R1, 99.99% U, 86.2% Pu and a part of Am or Eu are stripped into aqueous phase by 0.2mol/L acetohydroxyamic acid (AHA) in 0.01mol/L $HNO_3$ solution. In battery $R_2$, Am, Eu and remained Pu are completely back-extracted by 0.2mol/L AHA. This separation process contains no salt reagent, and it is not necessary to dilute HLLW feed.