Within the air purification system of a nuclear power plant, specific radioactive isotopes are extracted from gases through adsorption onto activated carbon. To properly dispose of used activated carbon, it is essential to determine the concentration of radioactive nuclides within it. This study discusses the application of the pyrolysis method for analyzing the concentrations of 3H and 14C in spent activated carbon. The pyrolysis was conducted using Raddec’s Pyrolyser, with adjustments made to parameters such as temperature profiles, airflow rates, sample quantities, and trapping solution volumes. The evaluation method for the pyrolysis of activated carbon to analyze 3H and 14C involved adding 3H and 14C sources to the activated carbon before use and subsequently assessing the recovery rates of the added sources in comparison to the analysis results.

A disposal system for spent nuclear fuel mainly divides into two parts; Engineered barriers include spent nuclear fuel, canister, buffer and backfill and natural barriers mean a host rock surrounding engineered barriers. If radionuclides released from a repository, they can migrate to the ecosystem. Sorption plays an important role in retarding the migration of released radionuclides. Hence, the safety assessment for the disposal of a spent nuclear fuel should consider the migration and retardation of radionuclides in geosphere. Distribution coefficient is one of input parameters for the safety assessment. In this work, distribution coefficients for crystalline rock as a natural barrier were collected and evaluated for the purpose of safety assessment for the deep geological disposal of a spent nuclear fuel. The radionuclides considered in this work are as follows; alkali and alkaline earth metals (Cs, Sr, Ba), lanthanides (Sm), actinides (Ac, Am, Cm, Np, Pa, Pu, Th U), transition elements (Nb, Ni, Pd, Tc, Zr), and others (C, Cl, I, Rn, Se, Sn). The sorption of radionuclides is influenced by various geochemical conditions such as pH/carbonates, redox potential, ionic strength, radionuclide concentration, kinds and amounts of minerals, and microbes. For the evaluation of distribution coefficients, the data from Sweden (SKB), Finland (Posiva), Switzerland (Nagra), and Japan (JAEA) were collected, analyzed, and the recommended distribution coefficients have been suggested.

The organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA) can enhance the radionuclides’ solubility and have the potential to induce the acceleration of radionuclides’ mobility to a far-field from the radioactive waste repository. Hence, it is essential to evaluate the effect of organic complexing agents on radionuclide solubility through experimental analysis under similar conditions to those at the radioactive waste disposal site. In this study, five radionuclides (cesium, cobalt, strontium, iodine, and uranium) and three organic complexing agents (EDTA, NTA, and ISA) were selected as model substances. To simulate environmental conditions, the groundwater was collected near the repository and applied for solubility experiments. The solubility experiments were carried out under various ranges of pHs (7, 9, 11, and 13), temperatures (10°C, 20°C, and 40°C), and concentrations of organic complexing agents (0, 10-5, 10-4, 10-3, and 10-2 M). Experimental results showed that the presence of organic complexing agents significantly increased the solubility of the radionuclides. Cobalt and strontium had high solubility enhancement factors, even at low concentrations of organic complexing agents. We also developed a support vector machine (SVM) model using some of the experimental data and validated it using the rest of the solubility data. The root mean square error (RMSE) in the training and validation sets was 0.012 and 0.016, respectively. The SVM model allowed us to estimate the solubility value under untested conditions (e.g., pH 12, temperature 30°C, ISA 5×10-4 M). Therefore, our experimental solubility data and the SVM model can be used to predict radionuclide solubility and solubility enhancement by organic complexing agents under various conditions.

As a result of various generation, transmutation, and decay schemes, a wide variety of radionuclides exist in the reactor prior to accident occurrence. Considering all of the radionuclides as the accident source term in an offsite consequence analysis will inevitably take up excessive computer resources and time. Calculation time can be reduced with minimal impact on the accuracy of the results by considering only the nuclides that have a significant effect on the calculation among the potential radioactive sources that may be released into the environment. In earlier studies related to offsite consequence analysis, it is shown that the principal criteria for the radionuclide screening applied are as follows; radionuclide inventory in the reactor, radioactive half-life, radionuclide release fraction to the environment, relative dose contribution of nuclides within a specific group, and radiobiological importance. As a result, it is confirmed that 54, 60, and 69 nuclides are applied to the risk assessment performed in WASH-1400, NUREG-1150, and SOARCA (State-of-the-Art Reactor Consequence Analyses) project in the United States, respectively. In addition, in this study, the technical consultations with domestic and foreign experts were carried out to confirm details on criteria and process for screening out radionuclides in offsite consequence analysis. In this paper, based on the literature survey and technical consulting, we derived the screening process of selecting a list of radionuclides to be considered in the offsite consequence analysis. The first step is to eliminate radionuclides with little core inventory (less than specific threshold) or very short half-lives. However, important decay products of radionuclides that have short half-lives should not be excluded by this process. The next step is to further eliminate radionuclides by considering contribution to offsite impact, which is defined as a product of radioactivity released to the environment (i.e. ‘inventory in the reactor’ times ‘release fraction to offsite’) and comprehensive dose (or risk) coefficient taking into account all exposure pathways to be included. The final step is to delete isotopes that contribute less than certain threshold to any important dose metric through additional computer runs for each important source term. Even though it is presumed that this process is applicable to existing light water reactors and the set of accidents that would be considered in PSA, some of the assumptions or specific recommendations may need to be reconsidered for other reactor types or set of accident categories.

A disposal system for spent nuclear fuel divides into two parts; (1) engineered barriers including spent nuclear fuel, canister, buffer and backfill, (2) natural barriers surrounding engineered barriers. Sorption and diffusion are main retardation mechanisms for the migration of released radionuclides. We analyzed the sorption properties of radionuclides for bentonite as a buffer material and collected/ evaluated the distribution coefficients for the purpose of safety assessment for the deep geological disposal of a spent nuclear fuel. Through this, we presented recommended distribution coefficients for radionuclides required for the safety assessment. This work included the radionuclides as follows; alkali and alkaline earth metals (Cs, Sr, Ba), lanthanides (Sm), actinides (Ac, Am, Cm, Np, Pa, Pu, Th U), transition elements (Nb, Ni, Pd, Tc, Zr), and others (C, Cl, I, Rn, Se, Sn). The sorption of radionuclides affected various geochemical conditions such as pH/carbonates, redox potential, ionic strength, radionuclide concentration, kinds and amounts of minerals, and microbes. Among the evaluated radionuclides, Cs, Ni, Pd, and Ra is sensitive to the ionic strength, while Np, Pu, U, Se, and, Tc is sensitive to the redox condition. For the evaluation of distribution coefficients, the data from Sweden (SKB), Finland (Posiva), Switzerland (Nagra), and Japan (JAEA) were collected, analyzed, and the recommended distribution coefficients were suggested.

In order to measure the Radionuclides and Concentration, the directly grinded land samples (river soil, pine leaves and mugwort) among the environment samples around the nuclear power plant were filled in a 450 mL Marinelli beaker and weighed to obtain the dry mass ratio of the samples. Then the background and land samples were measured for 80,000 sec. The analysis of the collected land samples showed that most of them contained less radiation nuclide than the detection minimum limit in the ‘Ministry of Education, Science and Technology Public Notice No. 2010-32.’In others, the natural radionuclides 40K were detected. Of the products of nuclear reaction discharged by a nuclear reaction, 134Cs and 137Cs are more easily detected, and their discharge sources can be traced using the relative ratio. Although the radioactive concentration in the vicinity of Kori Nuclear Power Plant, which is more than 1,100km away from Fukushima, the Japanese nuclear accident site, continuous monitoring is needed as the radionuclides can still be accumulated in the soil or animals and plants.