This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

As the design life of nuclear power plants are coming to the end, starting with Kori unit 1, nuclear power related organizations have been actively conducted research on the treatment of nuclear power plant decommissioning waste. In this study, among various types of radioactive waste, stabilization and volume reduction experiments were conducted on radioactive contaminated soil waste. Korea has no experience in decommissioning nuclear power plants, but a large amount of radioactively contaminated soil waste was generated during the decommissioning of the KAERI research reactor (TRIGA Mark- II) and the uranium conversion facility. This case shows the possibility of generating radioactive soil waste from nuclear power plants and nuclear-related facilities sites. Soil waste should be solidified, because its fluidity and dispersibility wastes specified in the notification of the Korea Nuclear Safety and Security Commission. In addition, the solidified waste forms should have sufficient mechanical strength and water resistance. Numerous minerals in the soil are components that can make glass and ceramics, for this reason, glass-ceramic sintered body can be made by appropriate heat and pressure. The sintering conditions of soil were optimized, in order to make better economical and more stable sintered body, some additives (such as additives for glass were mixed) with the soil and sintering experiments were conducted. Uncontaminated natural soil was collected and used for the experiment after air drying. Moisture content, pH, bulk density, and organic content were measured to understand the basic properties of soil, and physicochemical properties of the soil were identified by XRD, XRF, TG, and SEM-EDS analysis. In order to understand the distribution by particle size of the soil, it was divided into Sand (0.05–2 mm) and Fines (< 0.05 mm). The green body was manufactured in the form of a cylinder with a diameter of 13mm and a height of about 10mm. Appropriate pressure (> 150 MPa) was applied to the soil to make a green body, and appropriate heat (> 800°C) was applied to the sintered body to make a sintered body. The sintering was conducted in a muffle furnace in air conditions. The volume reduction and compressive strength of the sintered body for each condition were evaluated.

Concrete is one of the largest wastes, by volume, generated during the decommissioning of nuclear facilities, which significantly influences the projected costs for the disposal of decommissioning wastes. Concrete consists of aggregates and a cement binder. In radioactive concrete, the radioisotopes are mainly associated with the cement component. If the radioactive isotope can be separated from the concrete to below the clearance criteria, the volume of radioactive concrete waste could be reduced effectively. We were studied to separate the radioactive materials from the concrete by using the thermomechanical and chemical treatment processes, sequentially. From the study, separated aggregate could be treated to achieve the clearance level. However, these processes generate a large volume of secondary acidic radioactive wastewater, which might be a critical problem to reduce the volume of radioactive concrete waste. In this research, separating the 137Cs and 90Sr from dissolved concrete wastewater to below the discharge criteria by precipitation method, it would be released to the environment under industrial waste guidelines. The experiments were conducted to using a simulated radioactive wastewater, formed by the dissolution of concrete within HCl, which was spiking the 137Cs and 90Sr, respectively. In addition, we applied the chemical precipitation methods with wastewater, using ferrocyanide for 137Cs and BaSO4 coprecipitation for 90Sr. As a result, targeted radionuclides could be removed to the discharge level (137Cs: 0.05 Bq·ml−1, 90Sr: 0.02 Bq·ml−1) by precipitation method. Therefore, it could reduce the secondary wastewater effectively by precipitation method and enhance the additional volume reduction for radioactive concrete waste.

Radioactive carbon, C-14, can be generated by the neutron capture reaction of O-17 during the nuclear power plant operation. Since C-14 is classified as an intermediate level waste radionuclide, it is required that an effective separation process for C-14. C-14 is mainly absorbed on activated carbon in the air cleanup system. Therefore, the main generation source of C-14 during the nuclear power plant decommissioning is spent activated carbon. KAERI has been developing the treatment of spent activated carbon. In this process, C-14 can be desorbed as a gaseous oxide form from the spent activated carbon at high-temperature vacuum conditions. This radioactive carbon dioxide can be captured into alkaline earth metal incorporated glass and can be transformed into carbonate form. However, the carbonate (e.g. CaCO3 and SrCO3) is dispersive. When the radioactive carbonates are disposed into a geological repository, they should be immobilized to remove future uncertainty. This study examined the stabilization/immobilization of the radioactive carbonates by the cement hydration process. Cement wasteform incorporated with calcium carbonate and strontium carbonate was produced under various waste loading (e.g. 20wt%, 40wt%, and 60wt% of CaCO3 and SrCO3, respectively). Then we evaluated mechanical and chemical durability by measuring compressive strength and leachability according to standard test methods specified in the waste acceptance criteria of the Gyeongju low and intermediate level waste repository (WAC-SIL-2022-1). Also, microstructure and thermal characteristics were investigated by SEM-EDS and TGA analysis.

A GoldSim Total System Performance Assessment has been developed and utilized for assessment of the various conceptual HLW repositories for spent nuclear fuels during last a few decades. Even though, almost all required parameter values associated with the repository system are frequently assumed or sometimes overestimated, they are still far from being highly reliable. Uncertainties nested in nuclide transport modeling around the repository are mainly dominated by these parametric uncertainties aside from intrinsic model uncertainty. Reliable estimate of the parameter values commonly expressed as probability density functions (PDFs) always require a large amount of measured data. Such input distributions are used as input to the probabilistic assessment program through Monte Carlo simulation to quantitatively provide possible uncertainty of the results. However, in most cases, especially in the safety assessment of the repository which is typically related with both long-time span and wide modeling domain, inefficient observed data from the field measurements are common, making conventional probabilistic calculations rather even uncertain. Since Bayesian approach is known to be especially powerful and efficient in the case of lacking of available data measured, such short data could be compensated by coupling with a priori belief, reducing uncertainty. By allowing the a priori knowledge for incorporating insufficient observed data, which include expert’ elicitation, their beliefs and judgment regarding the parameters as well as recent site-specific measurements, based on the Bayes’ theorem, the older parameter distributions, “prior” distribution can be updated to a rather newer and reliable “posterior” distribution. Newer distributions are not necessarily expressed as PDFs for probabilistic calculation. These updates could be done even iteratively as many times as data values are sequentially available, which calls sequential Bayesian updating, making belief of posterior distributions become much higher by reducing parametric uncertainty. To show a possible way to enhance the belief as well as to reduce the uncertainty involved in parameter for the Bayesian scheme, nuclide travel length in the far-field area of a hypothetical deep borehole spent fuel Repository was investigated. The algorithm and module that have been developed and implemented in GSTSPA through current study was shown to work well for all assumed prior, three sequential posterior distributions and likelihoods.

To decrease area of the repository for high-level radioactive waste, enhancing the disposal efficiency is needed for public acceptance. Previous studies regarding the performance assessment of KRS and KRS+ repository did not consider area-based variations of the geothermal gradient and rock thermal properties in Korea. This research estimated deposition hole spacing based on performance assessment of a repository using the distribution of geothermal gradient and rock thermal properties in Korea to increase disposal efficiency. Distributions of geothermal gradient, rock thermal properties were investigated based on 2019 Korea geothermal atlas published by Korea Institute of Geoscience and Mineral Resources (KIGAM). Effect of thermal performance parameters was analyzed using coupled thermal-hydraulic numerical simulations, and effect of rock thermal conductivity and deposition hole spacing on the maximum temperature of buffer was relatively large. In addition, distribution maps of thermal performance of a repository and deposition hole spacing were plotted using thermal performance parameters-maximum temperature of buffer regression equations and GIS data given by KIGAM. In the regions showing the highest maximum temperature of buffer in Korea, required deposition hole spacings were 10.5 m, 10.0 m, 10.1 m, respectively for KJ-II, MX-80, and FEBEX bentonite cases, and thereby additional disposal area of 40%, 33.3%, and 34.7% were required compared to that of the KRS+ repository. On the other hand, high disposal efficiency can be obtained in the regions showing the low maximum temperature of bentonite buffer. The methodology provided in this research can be used as one of the references for the selection of domestic candidate repository sites. Additional mechanical performance analysis should be conducted using distributions of mechanical properties of rock mass in Korea.

The research for the safe management of high-level waste in Korea has been conducted by the Korea Atomic Energy Research Institute since 1997, and the results have formed the basis of the national basic plan for the high-level waste management and the revised national basic plan. In the future, it is evolving and developing R&D focusing on securing technologies for demonstration of the disposal technologies and R&D to develop disposal concepts that increase safety and improve efficiency. Efficient management of heat generated from high-level radioactive waste, including spent nuclear fuel, is an important factor in establishing the disposal concepts because it must be in harmony with key factors such as repository layout, waste disposal container specifications, and design and operation for the barriers of the disposal system. For safe and complete isolation of highlevel radioactive waste in the deep geology, the disposal systems that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. These disposal systems were based on low burn-up spent nuclear fuel characteristics generated in the early stages of nuclear power generation, and next, based on the high-level wastes from recycling process of the high burn-up spent nuclear fuels, and were the direct disposal systems for the high burn-up spent nuclear fuels. So, it is necessary to track and analyze the change process in the decay heat characteristics of the high-level waste to be disposed of in order to improve the disposal concept, which enhances the safety of disposal and the utilization of the national land. Therefore, in this paper, the process of change in decay heat of reference spent nuclear fuels for disposal applied to the disposal concepts from the initial stage of development of high-level waste disposal technology to the present in Korea is analyzed.

This presentation summarizes recent research on estimating the mechanical loading environment of spent nuclear fuel (SNF) during normal storage and transportation scenarios sponsored by the US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program. Normal conditions of truck, ship, and railroad transportation of SNF were studied with testing and numerical modeling to determine that the shock and vibration loads applied to SNF during transportation are not expected to challenge SNF cladding integrity or the fatigue life of cladding. The 30 cm package drop scenario was studied with experiments and modeling to determine that mechanical loads during a 30 cm SNF package drop scenario are only expected to challenge SNF cladding integrity under worstcase conditions at elevated temperatures. The SFWST program is currently preparing seismic shake table testing to record SNF mechanical loads in a dry storage earthquake scenario. This presentation summarizes the findings of the transportation and package drop research and details the progress made on the current seismic test.

Collagen is one of the most widely used biological materials in medical design. Collagen extracted from marine organisms can be a good biomaterial for tissue engineering applications due to its suitable properties. In this study, collagen is extracted from fish skin of Ctenopharyngodon Idella; then, the freeze drying method is used to design a porous scaffold. The scaffolds are modified with the chemical crosslinker N-(3-Dimethylaminopropyl)-N'-ethyl carbodiimide hydrochloride (EDC) to improve some of the overall properties. The extracted collagen samples are evaluated by various analyzes including cytotoxicity test, SDS-PAGE, FTIR, DSC, SEM, biodegradability and cell culture. The results of the SDS-PAGE study demonstrate well the protein patterns of the extracted collagen. The results show that cross-linking of collagen scaffold increases denaturation temperature and degradation time. The results of cytotoxicity show that the modified scaffolds have no toxicity. The cell adhesion study also shows that epithelial cells adhere well to the scaffold. Therefore, this method of chemical modification of collagen scaffold can improve the physical and biological properties. Overall, the modified collagen scaffold can be a promising candidate for tissue engineering applications.

Bulbophyllum auricomum Lindl. is a rare orchid and has flowers with an attractive fragrance. The present study investigated the tissue culture method for micropropagation. Capsules derived from artificial self-pollination were obtained for the best seed germination in MS basal medium. Plant growth regulators (1.0 mg·L-1 of BAP and 2.0 mg·L-1 of NAA) were affected by callus induction from subcultured pseudobulb explants. For the callus subculture, different natural plant extracts were tested in 11 treatment media. Among them, MS medium with 150 mL·L-1 of coconut water was generally effective in fresh weight (1.75 ± 0.08) and (3.01 ± 0.20) of callus proliferation and PLBs induction at 1 and 2 months, respectively, followed by an MS combination of 30 g·L-1 of banana and 20 g·L-1 of potato extract. The results of a comparative study of different MS mediums containing plant growth regulators with a natural extract combination and MS medium supplemented with natural extract only showed that MS medium supplemented with a combination of natural extracts (150 mL·L-1 of coconut water) and plant growth regulators (2.0 mg·L-1 of BAP and 1.0 mg·L-1 of NAA) obtained the highest shoot regeneration (3.37 ± 0.17) and (6.41 ± 0.68) after 1 month and 2 months of culturing, respectively.

Meat affects color and quality by metabolite concentrations. Meat produces metabolites, and metabolites are caused by a variety of causes. Meat also produces metabolites by oxidation, which is an inevitable chemical process that meat undergoes which is resulting information of various chemical compounds. Thus, the aim of this study was to profiling the change of metabolites of M. longissimus lumborum during the storage at 4°C. Instrumental color measurements were showed decreasing chroma value, redness and yellowness (P<0.05) during storage, while non-significance (P>0.05) changes found in lightness value. Above all, hue angle was highest at 21 d of storage (P<0.05). The lipid and protein oxidation of muscles was measured by TBARS value significantly increased (P<0.05), thiol and carbonyl groups were also increased significantly (P<0.05) during the display. Total 19 of 60 identified compounds appeared to have a significant difference by storage time (P<0.05). Hue angle had a significant correlation with specific metabolites such as carbon disulfide, 3-methyl-1-butanol, 2-ethyl-1-hexanol, lactic acid and palmitic acid (P<0.05). Results of the current study provide the conversion of volatile and non-volatile metabolites and their correlation with oxidative indicators for changes in meat quality during aerobic storage.