In the case of decommissioning of a nuclear power plant, it is expected that a significant amount of VLLW and LLW that need to be disposed of are also expected. Conventional reduction technology is a method of extracting or removing radionuclides from waste, but this project is being carried out for the purpose of obtaining a reduction effect through the development of a material that treats another radioactive waste using radioactive waste. In this paper, the technology of impregnating LiOH capable of adsorbing radiocarbon to the gas filter material manufactured from concrete and soil waste as raw materials and the radiocarbon removal performance were reviewed. In this study, a raw material of ceramic filter was prepared by mixing concrete and soil waste with a powder of 40 m or less, and after sintering at 1,250°C, 5wt% to 40wt% of LiOH is impregnated with a filter capable of adsorbing carbon dioxide. was prepared. The prepared filter used ICP-OES and XRD to confirm the LiOH deposition result, and the concentration of carbon dioxide discharged through the carbon dioxide adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed depending on the flow rate of carbon dioxide supplied and the amount of material. Through this, it was possible to confirm the possibility of power generation in the adsorption performance of gas. In this study, after crushing waste concrete and waste soil, powders of 40 m or less were mixed with other additives to prepare raw materials for ceramic filters, and sintered at 1,250°C to manufacture filters. 5wt% to 40wt% of LiOH was impregnated on the prepared filter to give functionality to enable carbon dioxide adsorption. The results of LiOH deposition were confirmed using ICP-OES and XRD, and the change in the concentration of carbon dioxide emitted through a separately prepared adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed according to the flow rate of carbon dioxide supplied and the amount of material, and the possibility of developing a material for radioactive waste treatment using radioactive waste was confirmed when the porosity and specific surface area of the filter material were increased.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

The 2-round Delphi survey and Focus Group Interview (FGI) survey method, in this study, are sequentially applied for the level analysis of the high-level radioactive waste (HLW) management technologies, that are classified into transport/storage, site evaluation, and disposal categories. The 2- round Delphi survey was conducted on domestic 56 experts in the HLW field in Korea, and survey answers were managed with questionnaires distributed by e-mail. In the FGI survey, domestic 24 experts from management field were formed into three groups to conduct in-depth interviews. Past research achievements including journal papers, intellectual properties and the expert opinions presented at expert hearing on HLW technology were used as reference materials. As a result of the survey, in this study, the average domestic technology level compared to the leading countries was 83.1% in transport area, 79.6% in storage area, 62.2% in site evaluation area, and 57.4% in disposal area, respectively. When compared to the former level analysis results in 2017, technology level of transport-storage area increased by 8.6%, and the site evaluation-disposal technology area decreased by 7.27%. The highest factor that increase the level of technology in the transport-storage field was due to the increased R&D program resulting on journal papers, intellectual properties. In addition, the decrease factor in the level of technology in the site evaluation-disposal field was mainly due to relatively low R&D program when compared to the leading countries. Suggested method for the level survey can be used to find out the basic data of the lower tech technologies, to estimate the efficient research budgets and to prepare the R&D human resources. With this regards, R&D roadmap can be matured with suggested prediction method for the domestic technology level on HLW.

Copper is used for deep geological disposal canisters of spent nuclear fuels, because of excellent corrosion resistance in an oxygen-free environment. However, sulfide formation during the long-term exposure under deep geological disposal condition can be harmful for the integrity of copper canisters. Sulfur around the canisters can diffuse along grain boundaries of copper, causing grain boundary embrittlement by the formation of copper sulfides at the grain boundaries. The development of copper alloys preventing the formation of copper sulfides along grain boundaries is essential for the longterm safety of copper canisters. In this research, the mechanisms of copper sulfide formation at the grain boundary are identified, and possible alloying elements to prevent the copper sulfide formation are searched through the first principle calculations of solute atom-vacancy binding energy and the molecular dynamics calculation of grain boundary segregation energy. The comparison with the experimental literature results on the mitigation of copper embrittlement confirmed that the theoretically identified mechanisms of copper sulfide formation and the selected alloy elements are valid. Thereafter, binary copper alloys were prepared by using a vacuum arc melting furnace. Sulfur was added during casting of the copper alloys to induce the sulfide formation. The cast alloys were cold-rolled into a plate after homogenization heat treatment. The microstructure and mechanical property of each alloy were investigated after recrystallization in a vacuum tube heat treatment furnace. The copper alloys developed in this study are expected to contribute in increasing the long-term safety of deep geological disposal copper canisters by reducing the embrittlement caused by the sulfide formation.

The analysis of uranium migration is crucial for the accurate safety assessment of high-level radioactive waste (HLW) repository. Previous studies showed that the migration of the uranium can be affected by various physical and chemical processes, such as groundwater flow, heat transfer, sorption/ desorption and, precipitation/dissolution. Therefore, a coupled Thermal-Hydrological-Chemical (THC) model is required to accurately simulate the uranium migration near the HLW repository. In this study, COMSOL-PHREEQC coupled model was used to simulate the uranium migration. In the model, groundwater flow, heat transfer, and non-reactive solute transport were calculated by COMSOL, and geo-chemical reaction was calculated by PHREEQC. Sorption was primarily considered as geo-chemical reaction in the model, using the concept of two-site protolysis nonelctrostatic surface complexation and cation exchange (2 SP NE SC/CE). A modified operator splitting method was used to couple the results of COMSOL and PHREEQC. Three benchmarks were done to assess the accuracy of the model: 1) 1D transport and cation exchange model, 2) cesium transport in the column experiment done by Steefel et al. (2002), and 3) the batch sorption experiment done by Fernandes et al. (2012), and Bradbury and Baeyens (2009). Three benchmark results showed reliable matching with results from the previous studies. After the validation, uranium 1D transport simulation on arbitrary porewater condition was conducted. From the results, the evolution of the uranium front with sequentially saturating sites was observed. Due to the limitation of operator splitting method, time step effect was observed, which caused the uranium to sorbed at further sites then it should. For further study, 3 main tasks were proposed. First, precipitation/ dissolution will be added to the reaction part. Second, multiphase flow will be considered instead of single phase Darcy flow. Last, the effect of redox potential will be considered.

The structural integrity of concrete silos is important from the perspective of long-term operation of radioactive waste repository. Recently, the application of acoustic emission (AE) is considered as a promising technology for the systematic real-time health monitoring of concrete-like brittle material. In this study, the characteristics of AE wave propagation through concrete silo of Gyeongju radioactive waste repository were evaluated under the effects of groundwater and temperature for the quantitative damage assessment. The attenuation coefficients and absolute energies of AE waves were measured for the temperature cases of 15, 45, 75°C under dry and saturated concrete specimens, which were manufactured based on the concrete mix same as that of Gyeongju concrete silo. The geometric spreading and material loss were taken into account with regard to the wave attenuation coefficient. The attenuation coefficient shows a decreasing pattern with temperature rise for both dry and saturated specimens. The AE waves in saturated condition attenuate faster than those in dry condition. It is found that the effect of water content has a greater impact on the wave attenuation than the temperature. The results from this study will be used as valuable information for estimating the quantitative damage at the location micro-cracks are generated rather than the AE sensor location.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

경주 방폐물 처분시설의 1단계 시설로 건설된 지하 사일로 구조는 2014년에 10만 드럼 규모로 완공되어 현재 운영중에 있다. 지하 사일로 구조는 지름 25m, 높이 50m로써 방폐물을 저장하는 실린더부분과 돔 부분으로 구성되어 있으며, 돔부분은 운영터널과 연결 되는 하부 돔 부분과 상부 돔 부분으로 구분할 수 있다. 지하 사일로 구조의 벽체는 철근콘크리트 라이너이고, 두께는 약 1m이다. 본 논문에서는 지하 사일로 구조의 건설과정 및 운영과정의 단계별 유한요소해석을 수행하였다. SMAP-3D 프로그램을 사용하여 2차원 축대칭 유한요소해석을 수행하였다. 2차원 축대칭 유한요소모델의 신뢰성을 검토하고자 3차원 유한요소해석도 수행하였다. 본 논문 에서는 지하 사일로 구조의 구조거동을 분석하고 구조적 안전성을 검토결과를 제시하였다.

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.

Sulfate-rich waste powder containing a radioactive nuclide is generated from chemical decontamination process and radioactive liquid waste treatment using ion exchange resin. The radioactive sulfate-rich waste powder should be stabilized for final disposal. The techniques for immobilization of the radioactive sulfate-rich waste powder such as hydraulic cement, geopolymer, and iron phosphate glass have been applied, however, there are limitation in these techniques. Firstly, the hydraulic cement cannot applied to the wastes containing high concentration of sulfate because the expansion, cracks, and disintegration can be happened in the waste form. Geopolymer has a low density although they can be used as a good binder. The iron phosphate glass can be utilized, however, a considerable amount of SO2 gas is emitted due to the high sintering temperature. In this study, immobilization of radioactive sulfate-rich waste powder was carried out to resolve above problems by applying low temperature sintering method using a low-melting glass. As a result, it was confirmed that the waste form has a high bulk density. The compressive strength of the waste form was over 40 MPa, which is higher than the acceptance criteria (≥ 3.44 MPa). From ANS 16.1 test, it was verified that the waste form met the acceptance criteria of the leachability index (≥ 6). It was also confirmed that the waste form was chemically durable through product consistency test (PCT). In addition, the chemical stabilities of waste forms were compared following the sintering condition and the composition of the waste forms. The difference of the chemical stability was explained by difference in the abundance of chemical form obtained from the sequential extraction test.

Decontamination of spent nuclear fuel from decommissioned nuclear reactors is crucial to reduce the volume of intermediate-level waste. Fuel cladding hulls are one of the important parts due to high radioactivity. Their decontamination could possibly enable reclassification as low-level waste. Fuel cladding hulls used in research reactors and being developed for conventional light water reactors are Al-Mg and Fe-Cr-Al alloys, respectively. Therefore, the recovery of these component metals after decontamination is necessary to reduce the volume of highly radioactive waste. Electrochemical approach is often chosen due to its simplicity and effectiveness. Non-aqueous solvents, such as molten salts (MSs) and ionic liquids (ILs), are preferred to aqueous solvents due to the absence of hydrogen evolution. However, MSs and ILs are limited by high temperature and high synthesis cost, along with toxicity issues. Deep eutectic solvents (DESs) are synthesized from a hydrogen bond acceptor (HBA) and donor (HBD) and exhibit outstanding metal salt solubility, wide electrochemical window, good biocompatibility, and economic production process. These characteristics make DES an attractive candidate solvent for economic, green, and efficient electrodeposition compared with aqueous solvents such acids or nonaqueous solvents such as MSs or ILs. In this research, the feasibility of electrodeposition of Al-Mg and Fe-Cr-Al alloys in ChCl:EG, the most common DES synthesized from choline chloride (ChCl) and ethylene glycol (EG), will be tested. A standard three-electrode electrochemical cell with an Au plated working electrode and Al wires for counter and reference electrodes is utilized. Two electrolyte solutions (Al-Mg and Fe-Cr-Al) are prepared by dissolving 100 mM of each anhydrous metal chloride salts (AlCl3, MgCl2, CrCl3, and FeCl2) in ChCl:EG. Cyclic voltammogram (CV) is measured at 5, 10, 15, and 20 mV·s−1 to observe the redox reactions occurring in the solutions. Electrodeposition of each alloy is performed via chronoamperometry at observed reduction potentials from CV measurements. The deposited surfaces and cross-sections are examined by scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS) to analyze the surface morphology, cross-section composition, and thickness. Authors anticipate that the presence of different metals will greatly affect the possibility of electrodeposition. It is expected that although all metals are distributed throughout the surface, the morphology, in terms of particle size and shape, would differ depending on metals. Different metals will be deposited by layers of an approximate thickness of a few μm each. This research will illustrate a potential for recovery and electrodeposition of other precious radioactive metals from DES.

As the decommissioning of nuclear power plants increases, there is an increasing interest in the amounts of radioactive waste. Especially, the radiation dose limit for packaging of radioactive wastes shall not exceed 2 mSv·h−1 and 0.1 mSv·h−1 on contact and at 2 m, respectively in South Korea. The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging plan, the radiation shielding was optimized and estimated in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

Starting with the permanent shutdown of Kori Unit 1, the first waste treatment facility in Korea will be built on the Kori site. In this facility, major process such as decontamination, cutting, radiation measurement and volume reduction of decommissioning waste are performed, and radioactive liquid waste is generated by the waste treatment process and personnel decontamination. The generated liquid waste is finally discharged to the sea through radioactive monitoring system after sufficient treatment to meet the standard radiological effluent control. Whereas the treated liquid waste is additionally diluted through the circulation water discharge conduit and discharged to the sea in the operating nuclear power plants, there is no circulation water in the waste treatment facility. Therefore, a new discharging method for dilution after treatment should be considered. In this paper, the treatment concept and discharge method of radioactive liquid waste system in waste treatment facility are reviewed.

Wolsong unit 1 (W1), which is a CANDU-6 type PHWRs that had been operated for 30 years since 1983, was shutdown in 2019. In this study, the radioactive waste levels of calandria and concrete structures were calculated to establish a decommissioning plan for W1. The specific systems within the scope of this study were grouped into 6 major categories as follows: Calandria, End Shield, Fuel Channel Assembly, Reactivity Control Device, End Shield Support, Vault. The main operating history of W1 is that the re-tubing project was performed. These characteristics and operation history were reflected in the evaluation. The neutron flux and energy spectrum of each structure were calculated by using MCNP code, and ORIGEN code is implemented to the calculation of radioactivity for each nuclide using the results from MCNP and the material information of the structure. As for the impurity information, ASTM B350, B351, B353 standard was used for zircaloy alloy. For other alloy, impurity information provided by NUREG/CR-3474 was applied. Since W1 is expected to be decommissioned immediately, the waste level was evaluated under cooling conditions for 5 years after permanent shutdown. Through the level evaluation of each component obtained as a result of the study, it can be used as basic data for the radioactive waste management of the decommissioning plan.

The type of accidents associated with the operation of a melting facility for radioactive metal waste is assumed to only marginally differ from those associated with similar activities in the conventional metal casting industry or the current waste melting facility. However, the radiological consequences from a mishap or a technical failure differ widely. Three critical and at the same time possible accidents were identified: (1) activity release due to vapor explosion, (2) activity release due to ladle breakthrough, (3) consequences of failure in the hot-cell or furnace chamber not possible to remedy using remote equipment.

Source localization technique using acoustic emission (AE) has been widely used to track the accurate location of the damaged structure. The principle of localization is based on signal velocity and the time difference of arrival (TDOF) obtained from different signals for the specific source. However, signal velocity changes depending on the frequency domain of signals. In addition, the TDOF is dependent on the signal threshold which affects the prediction accuracy. In this study, a convolutional neural network (CNN)-based approach is used to overcome the existing problem. The concrete block corresponding to 1.3×1.3×1.3 m size is prepared according to the mixing ratio of Wolseong low-to-intermediate level radioactive waste disposal concrete materials. The source is excited using an impact hammer, and signals were acquired through eight AE sensors attached to the concrete block and a multi-channel AE measurement system. The different signals for a specific source are time-synchronized to obtain TDOF information and are transformed into a time-frequency domain using continuous wavelet transform (CWT) for consideration of various frequencies. The developed CNN model is compared with the conventional TDOF-based method using the testing dataset. The result suggests that the CNN-based method can contribute to the improvement of localization performance.

In order to indirectly evaluate the inventory of difficult-to-measure (DTM) nuclides in radioactive waste, the scaling factor method by key nuclide has been used. It has been usually applied to low-and intermediate-level dry active waste (DAW), and the tolerance of 1,000% margin of error in the US, that is the factor of 10, is applied as an allowable confidence limits considering the inhomogeneity of the waste and the limitation of sample size. This is because the scaling factor method is based on economic efficiency. Confidence limits is the uncertainty (sampling error) according to predicting the mean value of the population by the mean value of the sample at 95% confidence level, reflecting the limitations of sample size (representation) with the standard deviation. If the standard deviation is large, the sample size can be increased to satisfy the allowable confidence limits. In the new nuclear power plants, the concentration of cesium nuclide (137Cs) in radioactive waste tends to be very low due to advances in nuclear fuel and reactor core management technology, which makes it very difficult to apply cesium as a key nuclide. In addition, it is inevitable to apply the mean activity concentration method, which reasonably and empirically derives the concentration of DTM nuclides regardless of key nuclide, when the correlation between key and DTM nuclides is not significant. The mean activity method is a methodology that applies the average concentration of a sample set to the entire population, and is similar to applying the average concentration ratio between key and DTM nuclides of a sample set to the population in the scaling factor method. Therefore, in this paper, the maximum acceptable uncertainty (confidence limits) at a reasonable level was studied when applying the mean activity concentration method by arithmetic mean unlike the scaling factor method which usually uses the geometric mean method. Several measures were proposed by applying mutatis mutandis the acceptable standard deviation in radiation measurement and the factor of 10 principle, etc., and the appropriateness was reviewed through case analysis.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

Low to intermediate radioactive waste disposal concrete structures are subjected to coupled hydromechanical conditions and the identification of structural damage is crucial to ensure safe long-term disposal. Different damage models for concrete and the surrounding rock can affect the damage characteristics of radioactive waste disposal structures. In this study, the effects of different rock damage models are applied to the hydro-mechanical-damage coupled structural analysis of the Wolseong Low and Intermediate Level Radioactive Waste Disposal Center silo. A two-dimensional model of the disposal silo was modeled using the finite element analysis software COMSOL and the Mazars’ damage model was applied to the silo concrete. The Mazars’ model parameters were obtained from uniaxial compression and tensile tests on cylindrical concrete specimens after 28 days of water curing and further 32 days of wet curing at 75°C). The COMSOL embedded Richards equation module was used to simulate hydraulic analysis. Structural loading due to waste disposal was applied at the bottom of the silo structure and the damage evolution characteristics were investigated. The non-linear mechanical rock behavior obtained from laboratory tests (Hoek-Brown criterion, resonant column test, Mazar’s damage model) and field tests (Goodman Jack) were input to assess the effects of different rock damage models. The results highlight the importance of structural damage consideration when assessing the long-term stability and safety of underground radioactive waste disposal structures under coupled hydro-mechanical conditions.

Near-surface disposal facility is more susceptible to intrusion than underground repository, resulting in more possible pathways for contaminant release. Alike human intrusion, animals (e.g. Ants, Moles, etc.) could intrude into the disposal site to excavate burrows, which could cause direct release of contaminants to biosphere. In this paper, animal intrusion is demonstrated using GoldSim’s commercial contaminant transport module and impact on the integrity of the near-surface disposal facility is evaluated in terms of fractional release rate of the contaminants. In this study, the near-surface disposal facility is modelled with a single concrete vault to contain radionuclide according to LLW concentration limit stated in NSSC notice No.2020-6. The release of contaminants is modelled to occur directly after the institutional control period, and the contaminants are mostly transported from the concrete vault to cover layers via diffusion. To produce mathematical model of the release of the contaminants due to animal intrusion, firstly, the fraction of burrow volume for each cover layer is calculated separately for each animal species, based on their maximum possible intrusion depth. In this study, fractions of burrow volume for ants and moles are calculated based on their maximum possible intrusion depths, where for ants is 2–3 m, and for moles is 0.1–0.135 m. Then, assuming that the contaminants are distributed homogeneously throughout each cover layers by diffusion, fraction of contaminants transported into the uppermost layer via excavation of the burrow is calculated for each layer based on burrow volume, and fraction of contaminants removed from the uppermost layer to the layers below via collapse of the burrow is also calculated based on the burrow volume. Lastly, the net transportation of contaminants into and out of the burrow via excavation and collapse, respectively, is calculated and demonstrated using direct transfer rate function of the GoldSim. Based on the simulated result, the maximum mass flux is too minor to cause a meaningful impact on the safety. The peak mass flux of the most sensitive radionuclide, I-129, is witnessed at around year 1,470, with a flux value of 5.36×10−6 g·yr−1. This minor release of the contaminants could be due to cover layers being much thicker than the maximum possible intrusion depth of the animals, preventing the animal intrusion into the deeper layers of higher radionuclide concentration. In future, this study can be used to provide a guidance and fundamental data for scenario development and safety evaluation of the near-surface disposal facility.