The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.

Even though a huge amount of spent nuclear fuels are accumulated at each nuclear power plant site in Korea, our government has not yet started to select a final disposal site, which might require more than several km2 surface area. According to the second national plan for the management of high-level radioactive waste, the reference geological disposal concept followed the Finnish concept based on KBS-3 type. However, the second national plan also mentioned that it was necessary to develop the technical alternatives. Considering the limited area of the Korean peninsula, the authors had developed an alternative disposal concepts for spent nuclear fuels in order to enhance the disposal density since 2021. Among ten disposal concepts shown in the literature published in 2000’s, we narrowed them to four concepts by international experiences and expert judgements. Assuming 10,000 t of CANDU spent nuclear fuels (SNF), we designed the engineered barriers for each alternative disposal concept. That is, using a KURT geological conditions, the engineered barrier systems (EBS) for the following four alternative concepts were proposed: ① mined deep borehole matrix, ② sub-seabed disposal, ③ deep borehole disposal, and ④ multi-level dispoal. The quantitative data of each design such as foot prints, safety factors, economical factors are produced from the conceptual designs of the engineered barriers. Five evaluation criteria (public acceptance, safety, cost, technology readiness level, environmental friendliness) were chosen for the comparison of alternatives, and supporting indicators that can be evaluated quantitatively were derived. The AHP with domestic experts was applied to the comparison of alternatives. The twolevel disposal was proposed as the most appropriate alternative for the enhancement of disposal efficiency by the experts. If perspectives changes, the other alternatives would be preferred. Three kinds of the two-level disposal of CANDU SNF were compared. It was decided to dispose of all the CANDU spent nuclear fuels into the disposal holes in the lower-level disposal tunnels because total footprint of the disposal system for CANDU SNF was much smaller than that for PWR SNF. Currently, we reviewed the performance criteria related to the disposal canister and the buffer and designed the EBS for CANDU SNF. With the design, safety assessment and cost estimates for the alternative disposal system will be carried out next year.

It is expected that around 576,000 bundles of CANDU spent nuclear fuels (SNF) will be generated from the four CANDU reactors located at the Wolsong site. The authors designed and proposed a reference disposal concept based on the KBS-3 type and KURT geological data in 2022. In addition, we have reviewed the literatures and selected four alternative disposal methods to develop the higherefficiency disposal concept than the reference concept since 2021. As known well, the most important safety functions of the geological disposal are containment and isolation, and the secondary function is retardation. A disposal canister covers the former, and buffer may do the latter. In this study, we design the engineered barrier systems for the four alternative concepts: (1) mined deep borehole matrix, (2) sub-seabed disposal, (3) deep borehole disposal, and (4) multi-level dispoal. Assuming total 10,000 tU of CANDU SNF, four different kinds of unit disposal module consisting of disposal canisters and compacted bentonite buffers are designed based on the technique currently available. Two alternative concepts, sub-seabed disposal and multi-level disposal, share the same unit module design with the reference concept in 2022. For all the alternative concepts, we assume that the density of the compacted buffer is 1.6 g/cm3. For the mined deep borehole matrix disposal, we introduce a disposal canister slightly modified from the Canadian NWMO canister with a capacity of 48 bundles. The thickness of a copper layer is changed to be 10 mm considering the long-term corrosion resistance. The buffer thickness around a disposal canister is 20 cm, and the diameter of a borehole is 100 cm. Two different kinds of buffer blocks are proposed for the easy handling of them. For the deep borehole disposal, a SiC-stainless steel canister is designed, and 63 bundles of CANDU SNF is emplaced in the canister. We expect that the SiC ceramic canister shows very excellent corrosion resistance and has a high thermal conductivity under the geological conditions. The deep borehole will be plugged with four layered sealing materials consisting of granite blocks, compacted bentonite, SiC ceramic, and concrete plugs.

PWR spent nuclear fuel generally showed an oxide film thickness of 100 um or more with a combustion rate of 45 MWD/MTU or higher, while CANDU spent nuclear fuel with an average combustion rate of about 7.8 MWD/MTU had few issues related to hydride corrosion. Even based on the actual power plant data, it is known that the thickness of the oxide film is 10 μm or less on the surface of the coating tube, and brittleness caused by hydride is shown from the thickness of the oxide film of about 80 μm, so it is not worth considering. However, since corrosion may be accelerated by lithium ions, lithium ions may be said to be a very important factor in controlling the hydro-chemical environment of heavy water. Lithium has a negative effect on the corrosion of zirconium alloys. However, since local below 5 ppb to prevent corrosion. maintained at a concentration between 0.35 and 0.55 ppm. Hydrogen is known to have a positive effect by suppressing radioactive decomposition of the coolant and suppressing cracks in nickelbased alloys. However, too much hydrogen can produce hydride in a pressure tube composed of Zr-2.5Nb, so DH (Disolved Hydrogen) maintains the range of 0.27–0.90 ppm. pH and conductivity are completely determined by lithium ions, and DH can be completely removed below 5 ppb to prevent corrosion. Therefore, for cladding corrosion simulation of the CANDU spent nuclear fuel, a hydrochemical of the equipment, not 310°C, and 14 uS·Cm−1 is targeted as conditions for corrosion acceleration. In addition, for acceleration, the temperature was set to 345°C (margin 10°C), which is the maximum accommodation range of the equipment, not 310°C.