1.Introduction

Waste treatment technology for the separation and immobilization of radioactive nuclides generated from the pyrochemical process, which is one of the most viable technologies for the recycling of used nuclear fuels using electrochemical methods, has been intensively studied to achieve the reduction of radioactive waste volume. The pyrochemical process at KAERI consists of several sequential steps, such as voloxidation, electrolytic reduction, electrorefining, electrowinning, and waste salt treatment processes [1-3]. The wastes arising from the pyrochemical process after the sufficient use of chloride salts as molten state electrolytes are LiCl waste salt and LiCl-KCl eutectic waste salt. The LiCl waste salt [4-6] is generated from the electrolytic reduction process and the LiCl-KCl eutectic waste salt [7,8] is generated from the electrorefining and electrowinning processes. In particular, the LiCl waste salt contains highly heat generative group II fission products such as Sr and Ba nuclides. Therefore, the fission products within the waste salt should be separated and concentrated in small volumes to reduce the volume of final waste. Finally, the group II fission products should be fabricated into durable waste forms sustainable for several thousands of years to be disposed in a final geological repository.

Against these backgrounds, the present study reports the separation efficiency of group II fission products in LiCl waste salt through the layer-melt crystallization method [4-6] using Sr and Ba nuclides as surrogate materials of group II fission products. The nuclide separation efficiency of group II fission products has been evaluated under various cooling conditions of the crystallization process. Meanwhile, to convert concentrated group II chloride into stable oxide form in consideration of the solidification process, conversion/distillation process has been conducted, where selective oxidation of group II nuclides proceeds by Li₂O oxidant and residual salts are removed by the vacuum distillation process. Finally, to immobilize separated group II nuclides, a preliminary solidification study has been conducted using SiO₂-B₂O₃-Al₂O₃ matrix [9], and high density glass based waste form has been fabricated under 50wt% waste loading of strontium oxide surrogate material.

2.Concept of LiCl Waste Salt Treatment

The concept for treatment of LiCl waste salt from the pyrochemical process of used nuclear fuel was designed to minimize the radioactive waste generation and fabricate a durable waste form. Fig. 1 shows the treatment flow of the waste salt from the pyrochemical process. The treatment flow has two main processes. One is the purification process, and the other is the waste solidification process. In the purification process, the group II nuclides are separated as a thermally stable compound from the waste salt, and the waste salt is recovered into a renewable form to recycle to the pyrochemical process. The separated group II nuclides are fabricated into a durable waste form in the waste solidification process. The separated group II nuclide form and salt recycling have a great effect on the waste solidification process because materials and conditions for the solidification process can be changed according to chemical and physical properties of the separated group II nuclide form and it may not be easy to fabricate a durable waste form when the separated group II nuclides contain some salt phase [2,10,11]. In addition, the salt recycling is closely associated with a minimization of radioactive waste generation for the pyrochemical process. This means that the purification process has a vital role in the treatment of the waste salt from the pyrochemical process. Therefore, it is essential to design a purification process that can separate almost all of the group II nuclides in the waste salt into a stable form minimizing the salt phase within it and maximizing the salt recycled from the waste salt.

For these reasons, the purification process, composed of crystallization and conversion/distillation, was designed. The crystallization is a process to recover pure LiCl from LiCl waste salt and concentrate group II nuclides in residual LiCl melt using a phase diagram of LiCl-BaCl₂-SrCl₂. Through the conversion/distillation, BaCl₂ and SrCl₂ concentrated in LiCl are converted into thermally stable oxide forms using Li₂O, and are then separated from LiCl. In general, group II nuclide oxides have melting temperatures above 2,000°C, and are compatible with glass composites. Because of those, the conversion of group II nuclides into an oxide form was determined. The separated group II nuclide oxides are fabricated into a glass waste form using an inorganic composite (xSiO₂-yAl₂O₃-zB₂O₃).

3.Separation of Group II Nuclides from LiCl Waste Salt

3.1.Concentration of Group II Nuclides by Layer Melt Crystallization

To separate group II nuclides in LiCl waste salt generated from the electrolytic reduction process in the pyrochemical process, the layer-melt crystallization method has been employed to purify LiCl waste salt in the molten salt state. Previous studies [4-6] revealed that the Cs and Sr/Ba nuclides in LiCl waste salt are purified using layermelt crystallization with a separation efficiency of over 90%. However, the current study focused on the separation efficiency of the group II nuclides (Sr and Ba) using layer-melt crystallization. Lab-scale layer-melt crystallization apparatus is shown in Fig. 2 (a) that is composed of a crystallization plate and thermocouples monitoring the crystallization process (see Fig. 2 (b)), where the crystallization plate is rectangular in shape with a thickness of 15 mm, 60 mm wide, and 200 mm height and made by Inconel alloy. The purified LiCl crystals are obtained below its melting temperature of around 610°C on the crystallization plates which is cooled by Ar coolant. In this study, to evaluate group II nuclide separation efficiency using layer- melt crystallization method, the crystallization experiments were conducted under various cooling conditions. Detailed experimental conditions are tabulated in Table 1. The flow of cooling gas was slowly increased according to the table 1. For example, in the Ex-1, the cooling rate was increased from 0 to 30 L/min within 90 minutes with the cooling intensity of 0.333 L/min, where cooling intensity indicates the increasing rate of gas flow. The concentration of Sr and Ba nuclides are set based on the properties of LiCl waste salt generated after electrolytic reduction process. 3kg-LiCl having 1.11wt% SrCl₂ (33.41 g) and 2.31wt% BaCl₂ (69.26 g) were melted at 700 °C in a furnace. LiCl crystals were recovered by layer-melt crystallization, where the crystallization plate is cooled by Ar coolant gas with an increasing mass flow from 0 L/min to 30 L/min within 90 min, 180 min, 360 min, and 540 min for Ex-1, Ex-2, Ex-3, and Ex-4, respectively. Resulting cooling intensities that indicate Ar coolant increasing rate were 0.333 L/min, 0.167, 0.083, and 0.056 L/min for Ex-1, Ex-2, Ex-3, and Ex-4, respectively. Separation efficiencies (%) were calculated by the relationship of (C_before-C_after)/C_beforex 100, where C_before and C_after are nuclides concentration in the LiCl salt before and after crystallization process, which is analyzed by Inductively Coupled Plasma Atomic Emission Spectrometer (ICP-AES) with a Perkin Elmer (Optima 4300DV).

Fig. 3 shows purified LiCl salt on the crystallization plate for each experiment and resulting separation efficiencies are shown in Fig. 4. The crystal growth flux indicating weight of crystal growth per unit time and unit area of crystallization plate also calculated in Fig 4. (a). In the case of Ex-1, high cooling intensity resulted in high crystal growth flux of 0.06 g/min·cm², leading to low group II nuclide separation efficiency below 90% for both of Sr and Ba nuclide. However, as the crystal growth flux decreases, the separation efficiency has been increased to over 95% under the 0.02 g/min·cm² crystal growth flux condition. Also, the separation efficiencies according to the crystal growth rate indicating length of crystal growth per unit time showed same tendency with crystal growth flux (see Fig. 4 (b)). Consequently, through the layer-melt crystallization experiment, it was found that the group II nuclides in the LiCl waste salt are effectively separated with a high separation efficiency of over 95% under controlled crystal growth.

3.2.Conversion of Group II Nuclides and LiCl Salt Distillation

To separate group II nuclide (Sr) concentrated in LiCl waste salt generated from the crystallization process, conversion of SrCl₂ in LiCl into an oxide form and LiCl distillation were performed. The conversion of SrCl₂ in LiCl molten salts was carried out in an electric furnace which is able to heat to 1,200°C. The alumina crucible, as shown in Fig. 5 was used for the reaction boat. The sample was prepared as a mixture of 36 g of LiCl, 4 g of SrCl₂ and 1.5 g of Li₂O. The reaction was performed at 650°C for 2 h in the electric furnace, and the sample was stirred at an interval of thirty minutes by using a quartz rod during the reaction. After the conversion, LiCl in the sample was distilled in the TG furnace system (see Fig. 5). The temperature and pressure for LiCl vaporization was fixed at 750°C and 0.5 Torr, respectively. After the LiCl distillation, residues, as products of the conversion, were characterized to observe a chemical conversion of SrCl₂ by using XRD patterns. Sr contents in LiCl recovered from the LiCl distillation were analyzed using ICP-AES.

Before performing the conversion of SrCl₂ using Li₂O, the Gibb’s free energy of the reaction was calculated by using HSC-Chemistry 5.1 software [12], and the results were shown in Table 2. According to Table 2, the Gibb’s free energy is increased with temperatures, and the increment is not significant. It is desirable to perform the conversion at lower temperatures because LiCl is a high volatility compound. Therefore, the operation temperature for the conversion was determined to be 650°C. The content of SrCl₂ in LiCl surrogate waste was adjusted at 10wt% considering the concentration of group II nuclides after the crystallization process. Li₂O was injected in an amount corresponding to 2 equivalents for SrCl₂. After the conversion, the alumina crucible containing the salt was cooled to room temperature. The strontium oxide existed in LiCl molten salts as a liquid phase because they are soluble in LiCl molten salt. All the LiCl was vaporized by distillation and the liquid phase of the oxide was crystallized during the distillation process. This phenomenon is similar to the crystallization of the solid solution in a solvent [13].

After the distillation, the residues were identified, and the results were shown in Fig. 6. According to Fig. 6, the residues were characterized as SrO, Li₂O, and Li₅AlO₄ phase. As had been expected in the Gibb’s free energy, SrCl₂ was effectively converted into SrO. Li₂O existed in the products because of excessive Li2CO3 injection. It is considered that Li₅AlO₄ might be generated due to the reaction with alumina crucible and residual Li₂O. This can be prevented by using a graphite crucible because graphite is not reactive with Li₂O.

To evaluate the conversion efficiency of SrCl₂ into SrO, the concentration of Sr in LiCl salt recovered from the vaporization process was measured by ICP-AES analysis. As a result, the concentration of Sr in the recovered LiCl salt were equal to about 0.05 of the initial concentrations of Sr in LiCl molten salt. This indicates that the conversion/ distillation processes are very effective in the separation of group II nuclide from the LiCl waste salt.

4.Solidification of Separated Group II Nuclides

To immobilize separated group II nuclides in durable waste form, a preliminary solidification study was conducted using SiO₂-Al₂O₃-B₂O₃ glass matrix system. The glass waste forms were fabricated using a SiO₂-Al₂O₃- B₂O₃ glass matrix and commercially available strontium oxide materials as a representative surrogate waste material of group II fission products. A glass matrix of SiO₂- Al₂O₃-B₂O₃ (weight ratio of SiO₂:Al₂O₃:B₂O₃=3:2:1) and strontium oxide with 4-different compositions as shown in Table 3 were mixed and ground into fine powders in mortar and tapped into a graphite crucible (Inner diameter=20 mm, Height=150 mm) with a cap on the top. The graphite crucible was placed in an electric furnace, which is later charged with the Ar gas to make an inert atmosphere before heating. For a vitrification of mixed oxides, the crucible was heated up to 1,450°C with a heating rate of 6°C/ min and maintained at 1,450°C for 4 hours. The pressure in the furnace is maintained below 900 torr by venting the system upon heating. After vitrification, the glass wasteforms were slowly cooled down to room temperature and separated from the graphite crucible.

The glass waste forms for the immobilization of group II fission products are fabricated with waste loading ranging from 0 to 50wt% SrO and a fixed ratio of SiO₂-Al₂O₃- B₂O₃ glass matrix. The fabricated glass waste forms are shown in Fig. 7. It was found that the glass matrix was not vitrified under the formulation of #1 in Table 3. However, in the case of 50wt% SrO waste loading, the mixture was successfully vitrified, as shown in Fig. 7, and did not show devitrification under slow cooling conditions. This is because SrO acts as a glass modifier such as CaO in silicate glasses. Bulk density of #6 waste form, measured by Archimedes principle, was about 3.09 g/cm³; areasonably high density as a final waste form. Further study about physical and chemical properties of waste form having group II surrogate fission products with #6 formulation is now under investigation to verify its feasibility as a waste form.

5.Conclusion

In this study, to separate and solidify highly heat-generative group II fission products in LiCl waste salt generated from the pyrochemical process, successive waste treatment flow composed of a separation and solidification process has been established. The separation process combined by a concentration process of group II nuclides using layer-melt crystallization and the conversion/distillation process to obtain group II nuclide as a stable oxide form revealed their feasibility in the treatment of group II nuclides from LiCl waste salt. Finally, to solidify separated group II nuclides of oxide forms, vitrification was conducted using SiO₂-B₂O₃-Al₂O₃ matrix system and waste form with high waste loading and density was obtained. Consequently, treatment flow for the separation and solidification of group II fission products has been established through crystallization, conversion/distillation, and vitrification processes, and it would be beneficial for the reduction of final waste volume. It could also be applicable to various areas such as RTG (Radioisotopic thermoelectric generator) due to the highly heat generative Sr nuclide within the waste form.