Engineered barriers (concrete and grout) in Low- and Intermediate-Level Waste (L/ILW) disposal facilities tend to degrade by groundwater or rainfall water over a long period of time. During the degradation process, radionuclides stored in the disposal facility might be released into the pore water, which can pass through the natural rock barriers (granite and sedimentary rock) and may reach the near-field and far-field. In this transportation, radionuclide might be sorbed onto the engineered and natural rock barriers. In addition, the organic complexing agent such as ethylenediaminetetraacetic acid (EDTA) and α-isosaccharinic acid (ISA), is also present in pore water, which may affect the sorption and mobility of radionuclide. In this study, the sorption and mobility of 90Sr under different conditions such as two pHs (7 and 13), different initial concentrations of organic complexing agents (from 10-5 M to 10-2 M), and solutions (groundwater, pore water, and rainfall water) were investigated in a batch system. The groundwater was collected at the L/ILW disposal facility located at Gyeongju in South Korea. The pore water and rainfall water were artificially made in the laboratory. The concrete, grout, granite, and sedimentary rock samples were collected from the same study sites from where the groundwater was collected. The rock samples were crushed to 53-150 micrometers and were characterized by XRD, XRF, SEM-EDS, BET, and zeta potential analyzer. 90Sr concentration was determined using liquid scintillation counting. The sorption of 90Sr was described by distribution coefficients (Kd) and sorption reduction factor (SRF). In the case of EDTA, the Kd values of 90Sr remained constant from 10-5 M to 10-3 M and tended to decrease at 10-2 M, while in case of ISA the Kd values decreased steadily as the concentration of ISA was increased from 10-5 M to 10-3 M; However, a sudden reduction in the Kd values were observed above 10-2 M. In comparison to EDTA, ISA gave a higher SRF of 90Sr. Therefore, from the above results, it can be concluded that the presence of ISA has a greater effect on the sorption and mobility of radionuclide in the solutions than EDTA, and the radionuclide may reach near- and far-field of the L/ILW disposal facility.

Treatment methods such as interim storage and immobilization are being considered to dispose of intermediate level waste (ILW), but some wastes that have been treated in the past may require repackaging. Re-packaging means to cover repackaging of waste that has already been packaged in a waste container and re-packaging is required for the following reasons: loss of shielding or containment, damage to external handling features, package out-of-specification, insufficient records and external policy. The re-packaging includes various methods such as non-intrusive treatment, overpacking of waste package, external treatment of waste container, repair waste container, injection of stabiliser, disassemble waste package, high temperature process, and dissolve waste package. The purpose of this paper is to evaluate the re-packaging possibility for various wastes by identifying the main repackaging methods among the above various re-packaging methods. 1) Disposal outside of the waste container is a viable technique for most packages, as coating with a portable spray gun for low dose rate packages or remotely using a robotic arm for high dose rate packages. 2) Waste container repair is divided into welding repair and patching of waste container according to the degree of damage. Weld repair and patching are important techniques that can be used to add additional shielding, repair damaged areas, and improve the integrity of lifting gears that may not be compliant. 3) In general, disassembly of waste packages has been applied to loose drummed waste. Packages and waste forms are physically disassembled, reduced in size, and placed in different new packages. For practical solution, grouted waste is repackaged by cutting using proprietary equipment such as diamond saws, wire saws, core drilling and rupture techniques. 4) High-temperature process involves cutting the waste package and placing the pieces in a hot bath of inorganic liquid or molten metal, and the process is applicable to all waste types. However, treatment of all gases produced, compliance with waste types and acceptance criteria. Finally, dissolving waste packages, which is generally considered impractical due to the variety of chemicals and radionuclides present in ILW, is a process that is easier to perform on raw ILW than conditioned waste. An example of waste being re-packaged is when old drummed waste is recovered from an old storage facility and the waste needs to be repackaged into a form that meets modern standards for interim storage and disposal.

The treatment of radioactive waste by melting has been mainly discussed with low-level waste (LLW). Considering that a large amount of waste in RV or RVI is intermediate-level waste (ILW), however, it is necessary to examine the possibility of treatment by melting of ILW. Different from LLW, melting of ILW with a high content of long-lived nuclides would lead to no free releasee, but has advantages in volume reduction, homogenization, and delay of release. In this paper, the possibility of melting as an alternative technology for the treatment of ILW in the future is reviewed by analyzing the benefits generated by melting ILW in the following aspects: 1) Similar to melting techniques of LLW, them of ILW are mostly based on well-known techniques, but it is necessary to review the feasibility of performing operations such as removal of solidified melt using remote equipment in abnormal situations such as loss of electricity. 2) It is necessary to specify radiation limits for the melting operation unless the ILW melting operation technique can guarantee that the risk of abnormal occurrence is very low. The main quantified radiation parameter is the ingot dose rate, which of 10 mSv/h is considered more reasonable. 3) Although the treatment of ILW by melting leads to a reduction in volume, the main characteristics of the waste still remain, and no waste can be disposed of for free release. Thus, the main potential benefits are improved long-term safety and reduced waste volume. 4) Reducing the surface-to-volume ratio of the molten material could reduce the amount of corrosive material per unit time and, consequently, increase long-term safety. Its effect on long-term safety is difficult to quantify precisely as it depends on several factors, such as the geometry of the original component or whether radionuclides were distributed on the surface of the original component or the induced radioactivity. 5) The volume reduction of ILW is estimated to be reduced by about 1/4 compared to the generated amount when assuming a disposal volume reduction factor of 3 and considering the dose reduction due to radioactive decay after long-term storage, however, due to the lack of knowledge about non-hazardous facility alternatives, it is difficult to evaluate cost-benefit. This is heavily influenced by both the final volume reduction and the potential to reduce the complexity of the repository’s technical barriers.