Chelating agents, such as EDTA, NTA, and citric acid, possess the capacity to establish complexes with radionuclides, potentially enhancing the migration of such radionuclides from the disposal sites. Hence, quantification of these chelating agents in radioactive wastes is required to ensure secure disposal protocols. The determination of chelating agents in radioactive wastes is mainly composed of two steps, e.g. extraction and detection. However, there are little information on the extraction of the chelators in various radioactive wastes. We endeavored to optimize the extraction conditions for citric acid (CA) found within concrete, a prevalent component in the context of dismantled waste materials. Given the inherent high solubility of CA in water, we applied an aliquot of deionized water to the concrete and conducted a one-hour ultrasonic leaching procedure to facilitate chelate extraction. Subsequently, following the preparation of the concrete leachate via vacuum filtration and centrifugation to yield a clarified solution, we quantified the concentration of CA within the solution using Ion Chromatography (IC). To enhance leaching efficiency, we examined the % recovery variation with respect to the pH of the leaching solution. The optimized extraction method will be applied to diverse chelating agents and radioactive waste categories.

During the operation of a nuclear power plant (NPP), corrosion products called CRUD (Chalk River unidentified deposit) accumulate on the surface of the primary system. The CRUD components of pressurized light water reactors or heavy water reactors, represented by (NixFe1-x)(FeyCr1-y)2O4, are composed of Fe3O4, NiFe2O4, FeCr2O4, NiCr2O4, etc. Radionuclide such as Co-60 are deposited within this CRUD, so the entire deposited material must be dissolved and removed for decontamination. Chemical decontamination has the advantage of being able to decontaminate a wide metal surface, but has the disadvantage of generating a large amount of secondary waste. Recently, chemical decontamination methods that add an electrodynamic process are being studied to overcome these shortcomings. This technology is a method of dissolving CRUD by applying an electric field in the anodic compartment of a cell separated by CEM. It is a method of accelerating CRUD dissolution by generating a large amount of hydrogen ions in the anodic compartment. Dissolved metal ions pass through the CEM (cation exchange membrane) and move to the cathodic compartment (pH > 12), where they are removed by adsorption or precipitation process. Therefore, the speciation characteristics between decontamination agent (oxalic acid) and metal ions are very important. In this study, we investigated the speciation characteristics of Fe(II), Ni(II), Co(II) - oxalate, which are important complex species in CRUD dissolution cells. The thermodynamic equilibrium constant for hydrolysis of each ion and of M(II)-oxalate were collected and speciation characteristics were analyzed using the MINEQL 5.0 program. From the speciation characteristics of M(II)-oxalate, effective radionuclide removal methods in an electrodynamic cell were considered.

A disposal of radioactive wastes is one of the urgent issues in worldwide. Considering upcoming plans for decommissioning of nuclear power plants, this problem is unavoidable and should be discussed very thoughtfully before long. There are variety of methods to deal with radioactive wastes, including Incineration process, conventional gasification process and plasma gasification process and so on. Among them, plasma gasification process is in the limelight due to its ecofriendly features and very large volume reduction effects. So, lots of countries like Japan, Taiwan, Russia, Bulgaria are already utilizing commercial plasma melting facilities and researching their own characteristics & disposal abilities and so on. Within the scope of this paper, I would like to introduce other countries current status of plasma melting facilities, and reach the conclusion on the directions to go for realistic radioactive wastes treatment.

Every engineering decision in radioactive waste management should be based on both technical and economic considerations. Especially, the management of low-level radioactive waste (LLW) is more critical on economic concerns, due to its long-term and continuous nature, which emphasizes the importance of economic analysis. In this study, economic factors for LLW management were discussed with appropriate engineering applications. Two major factors that should be taken into account when assessing economic expectations are the accuracy of the results and its proper balancing with ALARA philosophy (As Low As Reasonably Achievable). The accuracy of the results depends on the correct application of alternatives within a realistic framework of waste processing. This is because the LLW management process involves variables such as component type, physical dimensions, and the monetary value at the processing date. Two commonly used alternatives are the simplified lump sum present worth and levelized annual cost methods, which are based on annual and capital costs. However, these discussions on alternatives not only pertain to the time series value of operational costs but also to future technical advancements, which are crucial for engineers. As new research results on LLW treatment emerge, proper consideration and adoption should be given to technical cost management. As safety is the core value of the entire nuclear industry, the ALARA philosophy should also be considered in the cost management of LLW. The typical cost of exposure in man-rem has ranged from $1,000 to $20,000 over the past decades. However, with increasing concerns about health and international political threats, the cost of man-rem should be subject to stricter criteria, even the balancing of costs and safety concerns is much controverse issue. Throughout the study, the importance of incorporating proper engineering insights into the assessment of technical value for the financial management of LLW was discussed. However, it’s essential to remember that financial management should not be solely assessed based on the size of expenses but rather by evaluating the current financial status, the value of money at the time, and anticipated future costs, considering the specific context and timeframe.

Kori Unit 1 was permanently shut down in 2017 and is currently being prepared for decommissioning. Decommissioning waste generated during the decommissioning of a nuclear power plant has the characteristic of being generated in large quantities over a short period. Therefore, if proper management is not carried out, abnormal situations (i.e., unauthorized disposal, diversion, etc.) may occur. According to IAEA General Safety Report Part 6, radioactive waste shall be managed for all waste streams in decommissioning. This means ensuring that all waste streams are managed by the recorded inventory of all decommissioning waste and verifying that the recorded inventory is reasonable. The radioactive waste management has been managed in units such as mass and radioactivity. However, in the case of decommissioning waste, the amount is very large, so management by radioactivity is expected to have limitations. Therefore, in this study, a simple test was conducted to verify the decommissioning waste generated by a hypothetical scenario by mass. In this study, establish a scenario assuming various flows of decommissioning waste expected to be generated and calculate the expected inventory of decommissioning waste using Microsoft Excel. Specifically, using “Material Unaccounted For” (MUF), a material balance equation in IAEA Services Series 15, Nuclear Material Accounting Handbook, the error inventory was calculated as the difference between the physical inventory of decommissioning waste in the area and the ending inventory. We propose a simple test scenario to verify the flow of decommissioning waste by verifying that the error inventory reasonably matches the set allowable error. This study aims to verify the inventory of decommissioning waste using the material balance methodology used for nuclear material accounting. It is expected that the safety and reliability of the nuclear power plant decommissioning process can be secured by verifying that the total inventory of equipment before decommissioning and the inventory of remaining equipment and decommissioning waste after decommissioning are reasonably consistent.

Activated carbon (AC), extensively used across various industrial sectors, serves as a sponge for different types of gases due to its porous carbon material. These gases are attracted to the carbon substrate via van der Waals forces. In nuclear power plants, AC is commonly used to adsorb radioactive gases such as 86Kr and 134Xe, as well as radioiodine sources like 131I and 133I from gaseous effluents. Even if the adsorbed radioactive gases and radioiodine decay into non-radioactive elements, the spent AC still contains radioactive species with long half-lives, such as 3H (Tritium, T) and 14C (radiocarbon). Minimizing and separating waste that contains long-lived nuclides (e.g., 14C) are pivotal components of an efficient waste management approach. A challenging aspect of effectively managing disposed AC is to minimize long-lived radioactive substances by eliminating them. This paper explores and summarizes the technology used to remove pollutants (3H, 14C) trapped within the pores of Activated carbon through thermochemical vacuum and surface oxidation processes. By recycling and reusing spent Activated carbon, we anticipate a reduction in the volume of radioactive waste, leading to decreased disposal costs. Furthermore, this paper will contribute as a valuable reference in future studies, enhancing the understanding of vacuum thermal desorption and surface oxidation of used Activated carbon.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

With the importance of permanent disposal of high-level radioactive waste (HLW) generated in Korea, the deep geological disposal system based on the KBS-3 type is being developed. Since the deep geological repository must provide the long-term isolation of HLW from the surface environment and normal habitats for humans, plants, and animals, it is essential to assess the longterm performance of the disposal facility considering thermal-hydraulic-mechanical-chemical (TH- M-C) evolution. Decay heat dissipated from HLW contained in the canister causes an increase in temperature in the adjacent area. The requirement for the maximum temperature is established in consideration of the possibility of bentonite degradation. Therefore, when designing the repository, the temperature in the region of interest should be identified in detail through the thermal evolution assessment to ensure that the design requirement is satisfied. In the thermal evolution analysis, it is needed to evaluate the temperature distribution over the entire area of the disposal panel to consider the heat generated from both a single canister and adjacent canisters. Computational fluid dynamics (CFD) codes are widely used for detailed temperature analysis but are limited to simulating a wide range. Accordingly, in this study, we developed an analytical solution-based program for efficiently calculating the temperature distribution throughout the deposition panel, which is based on threedimensional heat conduction equations. The code developed can assess the temperature distribution of engineered and natural barrier systems. Principal parameters to be inputted are as follows: (a) geometry of the panel (e.g. width, length, height, spacing between canisters), (b) geometry of the canister (e.g. diameter, height), (c) thermal properties of bentonite and host-rock, (d) initial conditions (e.g. residual heat, temperature), and (e) time information (e.g. canister emplacement rate, time-interval, period). Through the calculation for the conceptual problem of a deposition panel capable of accommodating 900 (i.e. 30×30) canisters, it was confirmed that the program can adequately predict when and where the maximum temperature will occur. It is expected that the overall temperature distribution within the panel can be obtained by the evaluation of the entire region using this program reflecting the detailed design of the repository to be developed in the future. In addition, the thermal evolution analysis considering the influence of other canisters can be performed by applying the results as boundary conditions in the CFD analysis.

According to the second high-level radioactive waste management national basic plan announced in December 2021, the reference geological disposal concept for spent nuclear fuels (SNF) in Korea followed the Finnish concept based on KBS-3 type. Also, the basic plan required consideration of the development of the technical alternatives. Accordingly, Korea Atomic Energy Research Institute is conducting analyses of various alternative disposal concepts for spent nuclear fuels and is in the final selection stage of an alternative disposal concept. 10 disposal concepts including reference concept were considered for analysis in terms of disposal efficiency and safety. They were reference concept, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, space disposal, sub-sea bed disposal, long-term storage, deep horizontal borehole disposal, and ice-sheet disposal. Among them, first 4 concepts, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, were selected as candidate alternative disposal concepts by the evaluation of qualitative items. And then, by the evaluation of quantitative and qualitative items with specialists, multi-level disposal concept was being selected as a final alternative disposal concept. Design basis and performance requirements for designing alternative disposal systems were laid in the previous stage. Based on this, the design strategy and main design requirements were derived, and the engineered barrier system of a high-efficiency disposal concept was preliminary designed accordingly. In addition, as an alternative disposal concept, performance targets and related requirements were established to ensure that the high-efficiency repository system and its engineered barrier system components, such as disposal containers, buffer bentonites, and backfill perform the safety functions. Items that qualitatively describe safety functions, performance goals, and related requirements at this stage and items whose quantitative values are changed according to future test results will be determined and updated in the process of finalizing and specifically designing an alternative highefficiency disposal system.

The WRK (Waste Repository Korea bentonite) compacted bentonite medium has been considered as the appropriate buffer material in the Korean SNF (Spent nuclear fuel) repository site. In this study, hydraulic properties of the WRK compacted bentonite core (4.5 cm in diameter and 1.0 cm in length) as the buffer material were investigated in laboratory experiments. The porosity and the entry pressure of the water saturated core at different confining pressure conditions were measured. The average velocity of water flow in the WRK compacted bentonite core was calculated from results of the breakthrough curves of the CsI aqueous solution and the hydraulic conductivity of the core was also calculated from the continuous flow core experiments. Because various gases could be generated by continuous SNF fission, container corrosion and biochemical reactions in the repository site, the gas migration property in the WRK compacted bentonite core was also investigated in experiments. The gas permeability and the average of gas (H2) in the core at different water saturation conditions were measured. Laboratory experiments with the WRK Compacted bentonite core were performed under conditions simulating the DGR environment (confining pressure: 1.5- 20.0 MPa, injection pressure: 1.0-5.0 MPa, water saturation: 0-100%). The WRK Compacted bentonite core was saturated at various confining pressure conditions and the porosity ranged from 27.5% to 43.75% (average: 36.75%). The calculated hydraulic conductivity (K) of the core using experimental results was 8.69×10-11 cm/s. The gas permeability of the core when the water saturation 0~58 % was ranged of 19.81~3.43×10-16 m2, representing that the gas migration in the buffer depends directly on the water saturation degree of the buffer medium. The average gas velocity in the core at 58% of water saturation was 9.8×10-6 m/s, suggesting that the gas could migrate fast through the buffer medium in the SNF repository site. Identification of the hydraulic property for the buffer medium, acquired through these experimental measurements is very rare and is considered to have high academic values. Experimental results from this study were used as input parameter values for the numerical modeling to simulate the long-term gas migration in the buffer zone and to evaluate the feasibility of the buffer material, controlling the radionuclide-gas migration in the SNF repository site.

Long-term climate and surface environment changes can influence the geological subsurface environment evolution. In this context, a fluid flow pathway developing and connection possibility can be increased between the near-surface zone and deep depth underground. Thus, it is necessary to identify and prepare for the overall fluid flow at the entire geological system to minimize uncertainty on the spent nuclear fuel (SNF) disposal safety. The fluid flow outside the subsurface environment is initially penetrated through the surface and then the unsaturated area. Thus, the previously proved reports, POSIVA in Finland, suggested that sequential research about the fluid infiltration experiment (INEX) and the investigation is necessary. Characterizing the unsaturated zone can help predict changes and ensure the safety of SNFs according to geological long-term evolution. For example, the INEX test was conducted at the upper part of ONKALO, about 50 to 100 m depth, to understand the geochemical evolution of the groundwater through the unsaturated zone, to evaluate the main flow of groundwater that can approach the SNF disposal reservoir, and to estimate the decreasing progress of the buffering capacity along the pathway through the deep geological disposal. In the present study, a preliminary test was performed in the UNsaturated-zone In-situ Test (UNIT) facility near the KAERI underground research tunnel to design and establish a methodology for infiltration experiments consistent with the regional characteristics. The results represented the methodological application is possible for characterizing unsaturated-zone to perform infiltration experiments. The scale of the experiment will be expanded sequentially, and continuous research will be conducted for the next application.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.