The buffer is installed around the disposal canister, subjected to heating due to decay heat while simultaneously experiencing expansion influenced by groundwater inflow from the surrounding rock. The engineering barrier system for deep geological disposal require the evaluation of longterm evolution based on the verification of individual component performance and the interactions among components within the disposal environment. Thus, it is crucial to identify the thermalhydro- mechanical-chemical (THMC) processes of the buffer and assess its long- and short-term stability based on these interactions. Therefore, we conducted experimental evaluations of saturationswelling, dry heating, gas transport, and mineralogical alterations that the buffer may undergo in the heated-hydration environment. We simulated a 310 mm-thick buffer material in a cylindrical form, simulating the domestic disposal system concept of KRS+ (the improved KAERI reference disposal system for spent nuclear fuel), and subjected it to the disposal environment using heating cartridges and a hydration system. To monitor the thermal-hydro-mechanical behavior within the buffer material, load cells were installed in the hydration section, and both of thermal couples and relative humidity sensors were placed at regular intervals from the heat source. After 140 days of heating and hydration, we dismantled the experimental cell and conducted post-mortem analyses of the samples. In this post-mortem analysis, we performed functions of distance from the water contents, heat source, wet density, dry density, saturation, and X-Ray diffraction analysis (XRD). The results showed that after 140 days in the heated-hydration environment, the samples exhibited a significant decrease water contents and saturation near the heat source, along with very low wet and dry densities. XRD Quantitative Analysis did not indicate mineralogical changes. The findings from this study are expected to be useful for input parameters and THMC interaction assessments for the long-term stability evaluation of buffer in deep geological disposal.

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.