Carbonates are inorganic ligands that are abundant in natural groundwater. They strongly influence radionuclide mobility by forming strong complexes, thereby increasing solubility and reducing soil absorption rates. We characterized the spectroscopic properties of Am(III)-carbonate species using UV-Vis absorption and time-resolved laser-induced fluorescence spectroscopy. The deconvoluted absorption spectra of aqueous Am(CO3)2 − and Am(CO3)3 3− species were identified at redshifted positions with lower molar absorption coefficients compared to the absorption spectrum of aqua Am3+. The luminescence spectrum of Am(CO3)3 3− was red-shifted from 688 nm for Am3+ to 695 nm with enhanced intensity and an extended lifetime. Colloidal Am(III)-carbonate compounds exhibited absorption at approximately 506 nm but had non-luminescent properties. Slow formation of colloidal particles was monitored based on the absorption spectral changes over the sample aging time. The experimental results showed that the solubility of Am(III) in carbonate solutions was higher than the predicted values from the thermodynamic constants in OECD-NEA reviews. These results emphasize the importance of kinetic parameters as well as thermodynamic constants to predict radionuclide migration. The identified spectroscopic properties of Am(III)-carbonate species enable monitoring time-dependent species evolution in addition to determining the thermodynamics of Am(III) in carbonate systems.
The electrochemical behavior was investigated during the electrolysis of nickel oxide in LiCl-Li2O salt mixture at 650℃ by changing several components. The focus of this work is to improve anode design and shroud design to increase current densities. The tested components were ceramic anode shroud porosity, porosity size, anode geometry, anode material, and metallic porous anode shroud. The goal of these experiments was to optimize and improve the reduction process. The highest contributors to higher current densities were anode shroud porosity and anode geometry.
The solubility and species distribution of radionuclides in groundwater are essential data for the safety assessment of deep underground spent nuclear fuel (SNF) disposal systems. Americium is a major radionuclide responsible for the long-term radiotoxicity of SNF. In this study, the solubility of americium compounds was evaluated in synthetic groundwater (Syn- DB3), simulating groundwater from the DB3 site of the KAERI Underground Research Tunnel. Geochemical modeling was performed using the ThermoChimie_11a thermochemical database. Concentration of dissolved Am(III) in Syn-DB3 in the pH range of 6.4–10.5 was experimentally measured under over-saturation conditions by liquid scintillation counting over 70 d. The absorption spectra recorded for the same period suggest that Am(III) colloidal particles formed initially followed by rapid precipitation within 2 d. In the pH range of 7.5–10.5, the concentration of dissolved Am(III) converged to approximately 2×10−7 M over 70 d, which is comparable to that of the amorphous AmCO3OH(am) according to the modeling results. As the samples were aged for 70 d, a slow equilibrium process occurred between the solid and solution phases. There was no indication of transformation of the amorphous phase into the crystalline phase during the observation period.
Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.
Numerical modeling and scenario composition are needed to characterize the geological environment of the disposal site and analyze the long-term evolution of natural barriers. In this study, processes and features of the hydro-mechanical behavior of natural barriers were categorized and represented using the interrelation matrix proposed by SKB and Posiva. A hydro-mechanical coupled model was evaluated for analyzing stress field changes and fracture zone re-activation. The processes corresponding to long-term evolution and the hydro-mechanical mechanisms that may accompany critical processes were identified. Consequently, practical numerical methods could be considered for these geological engineering issues. A case study using a numerical method for the stability analysis of an underground disposal system was performed. Critical stress distribution regime problems were analyzed numerically by considering the strata’s movement. Another case focused on the equivalent continuum domain composition under the upscaling process in fractured rocks. Numerical methods and case studies were reviewed, confirming that an appropriate and optimized modeling technique is essential for studying the stress state and geological history of the Korean Peninsula. Considering the environments of potential disposal sites in Korea, selecting the optimal application method that effectively simulates fractured rocks should be prioritized.
Colloid Formation and Migration (CFM) project is being carried out within the Grimsel Test Site (GTS) Phase Ⅵ. Since 2008, the Korea Atomic Energy Research Institute (KAERI) has joined CFM to investigate the behavior of colloid-facilitated radionuclide transport in a generic Underground Research Laboratory (URL). The CFM project includes a long-term in-situ test (LIT) and an in-rock bentonite erosion test (i-BET) to assess the in-situ colloid-facilitated radionuclide transport through the bentonite erosion in the natural flow field. In the LIT experiment, radionuclide-containing compacted bentonite was equipped with a triple-packer system and then positioned at the borehole in the shear zone. It was observed that colloid transport was limited owing to the low swelling pressure and low hydraulic conductivity. Therefore, a postmortem analysis is being conducted to estimate the partial migration and diffusion of radionuclides. The i-BET experiment, that focuses more on bentonite erosion, was newly designed to assess colloid formation in another flow field. The i-BET experiment started with the placement of compacted bentonite rings in the double-packer system, and the hydraulic parameters and bentonite erosion have been monitored since December 2018.
Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.
The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly 3H (500–3,600 Bq·g−1), 14C (7.5–29 Bq·g−1), 55Fe (1.1– 7.1 Bq·g−1), 59Ni (0.60–1.0 Bq·g−1), 60Co (0.74–70 Bq·g−1), 63Ni (0.60–94 Bq·g−1), 90Sr (0.25–5.0 Bq·g−1), 137Cs (0.64–8.7 Bq·g−1), and 152Eu (0.19–2.9) Bq·g−1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g−1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.
The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the keff value, 0.95. The STAD canister had the lower keff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in keff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.
Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y−1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.
The International Atomic Energy Agency (IAEA) entails independent decision-making for the safety supervision of civil nuclear facilities. To evaluate and review the safety of nuclear facilities, the national regulatory body usually consults independent institutions or external committees. Technical Support Organizations (TSOs) include national laboratories, research institutions, and consulting organizations. Support from professional organizations in other countries may also be required occasionally. Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. Accordingly, China has continuously improved upon the construction of such a system by establishing the National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional safety supervision stations, the Nuclear and Radiation Safety Center (NSC), a nuclear safety expert committee, and the National Nuclear and Radiation Safety Supervision Technology R&D Base, which serves as the test, verification, and R&D platform for providing consultation and technical support. An R&D system, however, remains to be formed. Future endeavors must focus on improving the technical support capacity of these systems. As an enhancement from institutional independence to capability independence is necessary for ensuring the independence of China’s nuclear safety regulatory institution, its regulatory capacity must be improved in the future.
Because of the massive development of nuclear power plants in China in recent years, China is facing the challenge of radioactive waste disposal. China has established complete regulatory requirements for radioactive waste disposal, but it also has encountered problems and challenges in low-level radioactive waste disposal in terms of management, selection of disposal facility sites, and implementation of a site selection plan. Three low-level radioactive waste disposal facilities that have been operated in China are described, and their activity limits, locations, and capacities are also outlined. The connotations of “regional” and “centralized” disposal policies are discussed in light of the characteristics of the radioactive waste. The characteristics and advantages of the regional and centralized disposal policies are compared. It is concluded that the regional disposal policy adopted in 1992 can no longer meet the current disposal needs, and China should adopt a combination of the two disposal policies to solve the problem of radioactive waste disposal.
When decommissioning nuclear power plant (NPP), the first task performed is cost estimation. This is an important task in terms of securing adequate decommissioning funds and managing the schedule. Therefore, many countries and institutions are conducting continuous research and also developing and using many programs for cost estimation. However, the cost estimated for decommissioning an NPP typically differs from the actual cost incurred in its decommissioning. This is caused by insufficient experience in decommissioning NPPs or lack of decommissioning cost data. This uncertainty in cost estimation can be in general compensated for by applying a contingency. However, reflecting an appropriate standard for the contingency is also difficult. Therefore, in this study, data analysis was conducted based on the contingency guideline suggested by each institution and the actual cost of decommissioning the NPP. Subsequently, TLG Service, Inc.’s process, which recently suggested specific decommissioning costs, was matched with ISDC (International Structure for Decommissioning Costing)’s work breakdown structure (WBS). Based on the matching result, the guideline for applying the contingency for ISDC’s WBS Level 1 were presented. This study will be helpful in cost estimation by applying appropriate contingency guidelines in countries or institutions that have no experience in decommissioning NPPs.