Chul-Hyung Kang

EXPERIMENTAL ASSESSMENT OF NON-TREATED BENTONITE AS THE BUFFER MATERIAL OF A RADIOACTIVE WASTE REPOSITORY

The bentonite-based material being evaluated in several countries as potential barriers and seals for a nuclear waste disposal system is of mostly sodium type, whereas most bentonite available in Korea is known to be of calcium type. In order to investigate whether local Korean bentonite could be useful as a buffer or sealing material in an HLW repository system, raw bentonites sampled from the south-east area of Korea were examined in terms of their physicochemical properties such as surface area, CEC, swelling rate, and distribution coefficient. The diffusion behavior of some radionuclides of interest in compacted bentonite was also investigated. Considering that HLW generates decay heat over a long time, the thermal effect on the physicochemical properties of bentonite was also included. Four local samples were identified as Ca-bentonite through XRD and chemical analysis. Of the measured values of surface area, CEC and swelling rate of the local samples, Sample-A was found to have the greatest properties as the most likely candidate barrier material. The distribution coefficients of Cs-137, Sr-85, Co-60 and Am-241 for Sample-A sample were measured by the batch method. Sorption equilibrium was reached in around 8 to 10 days, but that of Sr was found to be reached earlier. Comparing the results of this study with the reference data, domestic bentonite was found to have a relatively high sorption ability. For the effect of varying concentration on sorption, the values of Kd peaked at 10−9–10−7 mol/l of radionuclide concentration. In XRD analysis, the (001) peak of Sample-A was fully collapsed above 200°C. The shoulder appearing at about 150°C in the DSC curve was found to be evidence that Sample-A is predominated by Ca-montmorillonite. The loss of swelling capacity and CEC of Sample-A started at about l00°C. The swelling data and the (001) peak intensity of the heat-treated sample showed that they were linearly interrelated. The measured Kd values of Co-60, Cs-137 and Am-241 for the samples heat-treated at various temperature showed that the domestic bentonite still retained sorption capacity below 100°C. In addition to such findings of thermal effects, it was found that the presence of calcium in bentonite may help to assure long-term stability under the expected thermo-hydro conditions. The Da values of Sr-85, Cs-137, Co-60, Am-241 and Cl-36 were measured to be 1.073×101, 6.705×10−1, 1.226×10−1, 1.310×10−2 and 9.490×101 μm2/sec, respectively, which could be arranged with the magnitude of their distribution coefficients, i.e. Cl>Sr>Cs>Co>Am. As the as-pressed density of bentonite increasing from 1.8 to 2.0 g/cm3, the Da-value of Cs-137 decreased by 25%. From the analyses of the diffusion mechanism of radionuclides in compacted bentonite, the surface diffusion due to the concentration gradients of radionuclide sorbed on the bentonite particles was found to be a dominating transport process of radionuclides in compacted bentonite with 1.8 g/cm3. Bases on these results, it was identified that domestic bentonite has potential as a chemical barrier material in a repository system. Some data obtained in the results could contribute to the engineering parameters to design a waste package and engineered barrier or to develop an appropriate disposal concept satisfying the safety requirements.

THE DEVELOPMENT OF A SAFETY ASSESSMENT APPROACH AND ITS IMPLICATION ON THE ADVANCED NUCLEAR FUEL CYCLE

The development of Advanced Nuclear Fuel Cycle (ANFC) technology is essential to meet the national mission for energy independence via a nuclear option in Korea. The action target is to develop environmentally friendly, cost-effective measures to reduce the burden of long term disposal. The proper scenarios regarding potential radionuclide release from a repository have been developed in this study based on the Advanced Korean Reference Disposal System (A-KRS). To predict safety for the various scenarios, a new assessment code based on the GoldSim software has also been developed. Deterministic analysis indicates an environmental benefit from the ANFC as long as the solid wastes from the ANFC act as a proper barrier.

The effect of the pseudo-colloids in the fractured porous media for safety assessment of a HLW repository: application of arbitrary function

In a deep-geological disposal facility, most of the pseudo-colloids are generating at the interface between a near-field and a far-field and the colloid transport is mostly taking place in the far-field region. This study has developed the methodology with the tabular result from a near-field model of the radionuclide transport as a boundary condition. This methodology can analyze the effect of pseudo-colloids to be interconnected the far-field region. This methodology is verified analytical solutions and will be a good tool to analyze the effect of pseudo-colloids in the far-field region at the deep-geological disposal environment.

Comparison of the amount of nuclides released from the spent fuel in contact with and without a compacted bentonite block

A spent LWR fuel specimen between Ca-bentonite blocks was leached in a simulated bentonite-saturated granitic water (SBGW) for 165 days and its results were compared with those of a specimen leached without a bentonite block. The amounts of Cs, Sb, Sr, Eu, Am, U and Pu released from a 4.3 mm thickness of a fuel pellet with a 50,400 MWD/MTU burn-up in the SBGW without a bentonite block were 2.2, 0.25, 0.15, 0.02, 0.01, 0.01 and ∼5 × 10−4% of their inventories, respectively. However, the amounts of nuclides released from the specimen between the 1.4 Mg/m3 bentonite blocks were decreased by three times at least. Moreover, the concentrations of the nuclides in the leachate were very low because most of them were retained in the bentonite blocks.

Preliminary Results for the Experimental Evaluation of a Radwaste Repository Near-Field Model

A series of experiments for a simplified system has been performed to evaluate a near-field model of radioactive waste repository developed by the University of California, Berkeley. In the experiments, iodine as nonsorbing species and cesium as sorbing species were employed. This paper presents the initial 200-day results of these experiments. Good agreement is found between the model and the experimental results. The results of this work will give basic ideas on the design of the near field of a repository.