With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

When aluminum is in an alkaline state, the aluminum oxide film surrounding aluminum is dissolved and moisture penetrates the exposed aluminum surface, causing corrosion of aluminum. At this time, hydrogen gas is generated and there is a risk of explosion due to the generated hydrogen gas. Aluminum radioactive waste is difficult to permanently dispose of because it does not meet the standards for the acquisition of low- and intermediate-level radioactive waste cave disposal facilities currently managed and operated by the Korea Nuclear Environment Corporation. However, because of this risk, it is necessary to study how to safely treat and dispose aluminum waste. In this study, overseas cases were investigated and analyzed to ensure the safety of aluminum waste disposal, and the current status of aluminum radioactive waste generated during decommissioning of the Korea Research Reactor 1&2 and a treatment plan to secure disposal suitability were presented. The process of removing a little remaining oxygen in molten steel during the reduction of iron oxide in the iron refining process is called deoxidation, and a representative material used for deoxidation is aluminum. In the case of metal melting decontamination, which is one of the decontamination processes of radioactive metal waste, a method of treating aluminum waste by using aluminum as a deoxidizer is proposed.

Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

Over the years, in the field of safety assessment of geological disposal system, system-level models have been widely employed, primarily due to considerations of computational efficiency and convenience. However, system-level models have their limitations when it comes to phenomenologically simulating the complex processes occurring within disposal systems, particularly when attempting to account for the coupled processes in the near-field. Therefore, this study investigates a machine learning-based methodology for incorporating phenomenological insights into system-level safety assessment models without compromising computational efficiency. The machine learning application targeted the calculation of waste degradation rates and the estimation of radionuclide flux around the deposition holes. To develop machine learning models for both degradation rates and radionuclide flux, key influencing factors or input parameters need to be identified. Subsequently, process models capable of computing degradation rates and radionuclide flux will be established. To facilitate the generation of machine learning data encompassing a wide range of input parameter combinations, Latin-hypercube sampling will be applied. Based on the predefined scenarios and input parameters, the machine learning models will generate time-series data for the degradation rates and radionuclide flux. The time-series data can subsequently be applied to the system-level safety assessment model as a time table format. The methodology presented in this study is expected to contribute to the enhancement of system-level safety assessment models when applied.

The operation time of a disposal repository is generally more than one hundred years except for the institutional control phase. The structural integrity of a repository can be regarded as one of the most important research issues from the perspective of a long-term performance assessment, which is closely related to the public acceptance with regard to the nuclear safety. The objective of this study is to suggest the methodology for quantitative evaluation of structural integrity in a nuclear waste repository based on the adaptive artificial intelligence (AI), fractal theory, and acoustic emission (AE) monitoring. Here, adaptive AI means that the advanced AI model trained additionally based on the expert’s decision, engineering & field scale tests, numerical studies etc. in addition to the lab. test. In the process of a methodology development, AE source location, wave attenuation, the maximum AE energy and crack type classification were subsequently studied from the various lab. tests and Mazars damage model. The developed methodology for structural integrity was also applied to engineering scale concrete block (1.3 m × 1.3 m × 1.3 m) by artificial crack generation using a plate jacking method (up to 30 MPa) in KURT (KAERI Underground Research Tunnel). The concrete recipe used in engineering scale test was same as that of Gyeongju low & intermediate level waste repository. From this study, the reliability for AE crack source location, crack type classification, and damage assessment increased and all the processes for the technology development were verified from the Korea Testing Laboratory (KTL) in 2022.

Dry active waste (DAW) contains substantial amount of cellulose related materials. The DAW are usually classified as low and/or very low-level waste. In Korea, three types of disposal facilities have been considered: silo, engineering barrier, and land-fill. Currently, only the silo type disposal facility is in operation. Around 27 thousand drums were disposed in silo. Massive amount of cement concrete is used in construction of silo. The ground waste, which flow through the concrete structure, shows higher pH than as it is. It is generally known that the pH of silo is ~12.47 in Korea, when considering construction material, filling material, and property of ground water. It is expected that the cellulose in DAW will be partially transformed to isosaccharinic acid (ISA). It is generally accepted that the ISA plays a negative role in safety analysis of disposal facility by stimulation of specific nuclides. Various factors affect the degradation of cellulose containing radioactive waste, such as degree of polymerization, pH of disposal condition, interaction between concrete structure and ground water, etc. In this paper, the disposal safety analysis of cellulose containing radioactive, usually paper, cotton, wood, etc., are studied. The degradation of cellulose with respect to degree of polymerization, pH of neighboring water, filling material of silo, etc. are reviewed. Based on the review results, it is reasonable to conclude that the substantial amount of DAW could be disposed in silo.

To prevent the release of radionuclides into the biosphere, disposal facilities for radioactive waste should be located to provide isolation from the accessible biosphere for tens of thousands to a million years after closure. During the period of interest, the constantly evolving natural environment and possible geological events of the site can cause disturbances to the containment function of the repository. Thus, for the long-term safety assessment of the repository, the possible long-term change of natural barrier should be considered. Due to the characteristics of radionuclides that transport mainly through the groundwater, understanding the long-term evolution of groundwater flow and geochemical properties is essential to assess the long-term changes in the natural barrier performance. The changes in characteristics of natural rocks and geological structures are one of the main factors that determine the hydrological and geochemical characteristics of the deep underground. In this study, we plan to develop a methodology to estimate these future geological evolutions in order to assess the possibility of hazardous events of the site that can affect hydrological or geochemical properties over the period of interest, and also in order to verify the change in the geological environment is within the safe performance range even after the period of interest. However, it is very unreliable to predict future changes in the natural environment because it is very heterogeneous, complex, and difficult to observe directly. For the preliminary study of the project, we reviewed cases of future evolution prediction researches with regard to the geological environment of disposal site and methods they applied to reduce the uncertainty of the prediction. The results will be used to establish basic data for future studies on the long-term evolution of hydraulic-mechanics performance of natural barrier and long-term evolution of geochemical performance around KURT site. In addition, it can contribute to construct long-term evolution scenario of the geological environment around future URL site.

The timescale of safety assessment for a geological disposal system is considered up to hundreds of thousands of years when the radionuclides in spent nuclear fuel decay to levels comparable to natural radioactivity. During this long period, a variety of climate changes are expected to occur, including variations in temperature and precipitation as well as long-term sea level changes and glacial cycles. These climate changes can either directly affect water balance components or indirectly affect water balance by altering terrain and vegetation that have an impact on water balance. Water balance is a significant element of safety assessment, because it affects the radionuclide transport via groundwater flow, which in turn affects the radiological risk to humans and other biotas. Therefore, it is important to understand the hydrologic response to climate changes for proving the long-term safety of the disposal system. To this end, this study performed hydrological simulations using the SWAT (Soil and Water Assessment Tool) for several climate change scenarios. SWAT is the watershed-scale hydrological model developed by the USDA-ARS (United States Department of Agriculture - Agricultural Research Service) and has been widely used to quantify the water balance in a watershed. It calculates the hydrologic cycle based on the water balance equation with different physical processes for water balance components such as evapotranspiration, surface runoff, and groundwater recharge. This study assumed several climate change scenarios (e.g., variations in temperature and precipitation, sea level change, and formation of permafrost) and analyzed how the components of the water balance would respond under different scenarios and which scenarios would have the greatest impact on the water balance. These findings can provide valuable insights for future long-term safety assessments on the Korean Peninsula and can also be used as input data for the biosphere module of APro (Adaptive process-based total system performance assessment framework).

Since high-level radioactive wastes contain long-lived nuclides and emit high energy, they should be disposed of permanently through a deep geological disposal system. In Korea, the first (2016.07) and the second (2021.12) basic plans for the management of high-level disposal systems were proposed to select sites for deep geological disposal facilities and to implement business strategies. Leading countries such as Finland, Sweden and France have developed and applied safety cases to verify the safety of deep geological disposal systems. By examining the regulatory status of foreign leading countries, we analyze the safety cases ranging from the site selection stage of the deep geological disposal system to the securing of the permanent disposal system to the investigation, analysis, evaluation, design, construction, operation, and closure. Based on this analysis, we will develop safety case elements for long-term safety of deep geological disposal systems suitable for domestic situation. To systemically analyze data based on safety cases, we have established a database of deep geological disposal system regulations in leading foreign countries. Artificial intelligence text mining and data visualization techniques are used to provide database in dashboard form rather than simple lists of data items, which is a limitation of existing methods. This allows regulatory developers to understand information more quickly and intuitively and provide a convenient interface so that anyone can easily access the analyzed data and create meaningful information. Furthermore, based on the accumulated bigdata, the artificial intelligence learns and analyzes the information in the database through deep learning, and aims to derive a more accurate safety case. Based on these technologies, this study analyzed the legal systems, regulatory standards, and cases of major international leading countries and international organizations such as the United States, Sweden, Finland, Canada, Switzerland, and the IAEA to establish a database management system. To establish a safety regulation base suitable for the domestic deep geological disposal environment, the database is provided as data to refer to and apply systematic information management on regulatory standards and regulatory cases of overseas leading countries, and it is expected that it will play a key role as a forum for understanding and discussing the level of safety of deep geological disposal system among stakeholders.

Two sets of analyses for the cases of groundwater release to well and sea ecosystems were conducted for the environmental impact assessment of high-level radioactive waste disposal facilities. After obtaining the respective BDCF (Biosphere Dose Conversion Factor) results for the scenarios of well-farming and marine water fishing using different biosphere assessment conceptual models implemented in ECOLEGO, they were compared each other. The purposes of these analyses are to identify reference generic biosphere conceptual models and to get insight on model uncertainty. In this study, the endpoint used for the comparison of the ECOLEGO biosphere models was the socalled Biosphere Dose Conversion Factor (BDCF), which is defined as the maximum value of the total dose to the exposed group, in Sv/yr, resulting from a continuous unit release of 1 Bq/yr during the whole simulation time either to the well compartment (BDCF_Well) or to the marine water compartment (BDCF_Sea). The radionuclides considered in the comparison were Cs-137, I-129, Nb-94, Ni-59, Ni- 63, Sr-90 and Tc-99. The conceptual models used in the biosphere assessment of the releases to a well are based on models that have been used by the DOE (simple-soil model) and SKB (complex-soil model) in safety assessments of radioactive waste repositories, respectively. Difference between two conceptual models used in the assessment of the releases to a sea is the number of compartments representing the sea; i.e., one model represents the sea with one compartment for the water and one for the sediment (singlecompartment model), whereas the alternative model uses two compartments for the water and the sediments: one for the inner coast and one for the outer coast (double-compartment model). The results of the BDCF_Well to a farmer obtained with the DOE and SKB models are shown to be very close to each other. Despite the differences in conceptual models and parameters, the results are within a maximum difference of a factor of 4. The results from the SKB model were higher for all radionuclides. The values of the BDCF_Sea obtained with the single- and double-compartment models are shown to be larger differences with a maximum order of 2. For all studied radionuclides, the double-compartment model produces higher BDCFs than does the single-compartment model. The differences would be due to activity concentrations in both water and sediments. Since the hydrodynamic behavior assumed for flow in the sea could significantly influence the dilution volumes and hence the concentrations, it is found that site-specific investigations are necessary to establish an appropriate marine biosphere conceptual model.

The deep geologic repository (DGR) concept is widely accepted as the most feasible option for the final disposal of spent nuclear fuels. In this concept, a series of engineered and natural barrier systems are combined to safely store spent nuclear fuel and to isolate it from the biosphere for a practically indefinite period of time. Due to the extremely long lifetime of the DGR, the performance of the DGR replies especially on the natural geologic barriers. Assessing the safety of the DGR is thus required to evaluate the impacts of a wide range of geological, hydrogeological, and physicochemical processes including rare geological events as well as present water cycles and deep groundwater flow systems. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the numerical models used for safety evaluation need to be comprehensive, robust, and efficient. This study describes the development of an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be approved with confidence by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM under development can currently simulate overland flow, groundwater flow, near surface evapotranspiration in a modular manner. The IHM can also be considered as a framework as it can easily accommodate additional processes and requirements for the future as it is necessary. The IHM is capable of handling the atmospheric, land surface, and subsurface processes for simultaneously analyzing the regional groundwater driving force and deep subsurface flow, and repository scale safety features, providing an ultimate basis for seamless safety assessment in the DGR program. The applicability of the IHM to the DGR safety assessment is demonstrated using illustrative examples.

A radioactive waste disposal facility needs to be developed in a way to protect present and future generations and its environment. A safety assessment is implemented for normal and abnormal scenarios and human intrusion scenarios as a part of a safety case in developing a disposal facility for the radioactive waste. The human intrusion scenarios include a well scenario which takes into account various potential exposure groups (PEGs) who use a groundwater well contaminated with radionuclides released from the disposal facility. It is observed that a pumping rate has a negative correlation with the biosphere dose conversion factor (BDCF) in the well scenario. C-14 is shown to be a key radionuclide in the well scenario, and a special model based on the carbon cycle is applied for C-14. For Tc-99, an adsorption coefficient should be adjusted to be suitable for the site. The safety assessment for the radioactive waste disposal facility is successfully carried out for the well scenario. However, it is observed that site-specific models needs to be developed and sitespecific input data need to be collected in order to avoid unnecessary conservatism.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

To obtain confidence in the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. Thus, safety assessment models require uncertainty management as a key part of the confidencebuilding process. In application to the numerical modelling, the global sensitivity analysis is widely employed for dealing with parametric and conceptual uncertainties. In particular, the parametric uncertainty can be effectively reduced by minimizing the uncertainty of critical parameters in the safety assessment model. In this paper, the numerical model of each step disposal facility (Silo, Near surface, and Trench type) at Wolsong Low and Immediate Level Waste (LILW) Disposal Center is designed by using a two-dimensional finite element code (COMSOL Multiphysics). In order to determine the critical parameters for non-adsorbed nuclides such as H-3, C-14, Tc-99, we introduced the variance-based sensitivity analysis methodology of the global sensitivity analysis. In the case of Silo type, the density of waste is highly sensitive to the total leakage quantity of all nuclides. Additionally, the initial nuclide concentration of H-3 was identified as another important parameter of H-3. On the other hands, the mass transport coefficient showed a high contribution in C-14 and Tc-99. In other types of disposal facilities, the leaking properties of H-3 are significantly affected by the amount of infiltration water. However, C-14 and Tc-99 were found to be more sensitive to the density of waste.

Safety for the radioactive waste disposed of in the disposal facility should be secured through safety assessment in consideration of the various situations. In this study, the influence and correlation of EDTA and ISA, which are the factors that can impede the safety of the disposal facility, were analyzed using the PHREEQC computational code. Thermodynamic database (TDB) of Andra, specific ion interaction theory (SIT) model as ionic strength correction model, radionuclides (Ni, Am, Pu) were adopted to perform the calculation on the distribution of chemical species by pH. According to the results, EDTA dominated the system and the effect of ISA is relatively small for the distribution of the chemical species of divalent and trivalent cations in neutral and weak base conditions. In the case of the tetravalent cations, the effect of ISA increased compared to the previous case especially in the strong base conditions. In conclusion, EDTA has a more significant effect on the system than ISA under the environment of the domestic disposal facility. Furthermore, when EDTA and ISA are present simultaneously in the system, the effects of two materials are inversely proportional and this characteristic should be considered during the safety assessment.

In a recent preliminary inspection for disposal, the glass fiber waste (GFW), used as a pipe insulation, was judged as “pending evaluation” because some dust was found in drum opening tests. Therefore, additional inspection is required to ensure that the package corresponds with the acceptance criteria of the particulates. The dust was generated presumably due to GFW being used in a high-temperature environment for a long time, thus being easily degraded and crushed. For this reason, safety issues that may occur in the process of handling, transportation, and disposal are emerging. Therefore, in this study, a preliminary safety assessment of GFW disposal was performed, the exposure dose to the general public was derived, and compared with the dose limit. The evaluation was carried out in the following order: (1) evaluation of GFW radiation source term, (2) selection of accident scenario, (3) calculation of exposure dose, (4) comparison of evaluation results with dose limits, and confirmation of satisfaction. The average radioactivity of the GFW to be disposed of was used as the source term, and the main nuclides were identified as H-3, Fe-55, Co-60, Ni-63, and Pu-241. In general, the types of accidents that can occur at disposal facilities can be classified into falls, fires, collisions during transportation, off-site accidents, and nuclear criticality, and the accident scenarios are selected by analyzing and reviewing the probability of each accident. In this study, the accident analysis and scenarios presented in the safety assessment of the KORAD were reviewed, and the fire in the treatment facility, the fire in the storage facility, and the collision of the transport vehicle were selected as the evaluation scenarios. When an accident occurs, the radioactive material inside the container leaks out and diffuses into the atmosphere. In this evaluation, the internal and external exposure of the general public due to radioactive plume at the site boundary was evaluated and the dose conversion factors from ICRP-72 and FGR 12 were used. Based on the evaluation, general public was exposed to 0.004 mSv, 0.013 mSv, and 0.045 mSv, respectively, due to a fire at a treatment facility, at a storage facility, and in a transport vehicle. Most of the dose is due to internal exposure by Pu-241 nuclide, because the proportion of it in the waste is high, and when inhaled, the internal dose is high by emitting beta rays. It was confirmed that the result of dose was 0.4%, 1.3% and 4.5% of the annual dose limit, sufficiently satisfying the dose limit and safety.

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, the maximum safe disposal amount cellulose is evaluated considering the disposal environment of silos of 1st phase disposal facility. The key factor governing the impact of cellulose degradation is pH of disposal environment, as cellulose is known to degrade partially at pH above 12.5, and completely at pH above 13. Thus, disposal environment should be analyzed as to determine the extent of degradation. As silos are constructed with large amount of cement, porewater within concrete walls would be of very high pH. However, for high pH porewater to be released into the pores of crushed rock, which is filling up the silos, lower pH groundwater (commonly pH 7) should flow into the silos through the concrete walls. This causes dilution of the high pH concrete porewater, resulting in a lower pH as the silos are filled, reaching to expected pH of 11.8–12.3, which is below cellulose degradation condition. Thus, cellulose degradation is not expected, but to quantitatively evaluate safe disposal amount of cellulose, partial degradation is assumed. Upon literature review, the most conservative ISA concentration, enhancing radionuclide mobility, is determined to be 1.0×10−4 M and to reach this concentration, cellulose mass equivalent to 6wt% of cement of the repository, is required to be degraded. However, this ratio is derived based on complete degradation of cellulose into ISA, so for partial degradation, degradation ratio and yield ratio of ISA should be considered. Commonly, cellulosic material (e.g. cotton, paper, etc.) has degree of polymerization (DP) between 1,000–2,000, and with this DP, degradation ratio is estimated to be about 10%. Furthermore, yield ratio of ISA is known to be 80%. Considering all these aspects, about 1.79×107 kg of cellulose could be disposed, which if converted into number of drums, considering cellulose content of dry active waste, more than 100,000 drums (200 L) could be disposed with negligible impact on safety. Based on the result, negligible impact of cellulose degradation is expected for safety of 1st phase disposal facility. In future, this study could be used as fundamental data for revising waste acceptance criteria.

An objective of a safety assessment for geological disposal is to evaluate the radiological impact by radionuclides release from radioactive wastes. Computational estimation of all radionuclides transport in the disposal system, however, is not neccessary because some radionuclides has negligible effect on radiological doses. For this reason, prioritization of radionuclides list is preceded before the safety assessment. The Korea Atomic Energy Research Institue (KAERI) has assessed the long-term safety of a disposal system for spent nculear fuels. Currently, thirty eight radionuclides and twenty three elements are considered in the safety assessment activity of the KAERI. Nevertheless, a screening process for radionulides selection has not been articulated yet. In this study, we reviewed radionuclides selection process in forign countries to re-establish screening criteria for the KAERI’s radionuclides list. Screeing models of the Swedish Nuclear Fuel and Waste Management Company (SKB), the Deparment of Eenrgy (US DOE), and the Japan Nuclear Cycle Development Istitute (JNC) were compared. We found that each country developed different screening model depending on scenarios of radionuclides release. Nonetheless, there were common properties that determines the importance of radionuclides. These properties for radionuclides include halflife, radiotoxicity (or specific activity), and mobility in underground medium. Based on the review results, we proposed radionuclides selection process to prioritize the importance of radionucldies in the KAERI safety assessment.

Disposal facilities for radioactive waste shall be sited to provide isolation from the accessible biosphere. The features shall aim to provide this isolation for tens of thousands to a million years after closure. For the safety assessments of repository, the long-term natural evolution and possible events of the site, that can cause disturbances to the facility over the period of interest, should be considered. Geological development processes that the site have been experienced can contribute to understanding and descripting the present-day conditions. Moreover, knowledge of the past is necessary to predict the future evolution of the site. With regard to disposal site, understanding past geological evolution history allows to access the possibility of hazardous events of the site that can cause disturbances to the facility over the period of interest, and to verify the change in the geological environment is within the safe performance range even after the period of interest. In addition, certain parameters that change with the geological evolution can affect the hydrological and geochemical characteristics which are essential to disposal performance. There are various factors in the evolution of the geological environment, but not all are related to disposal safety. The objective of this research is to develop a geological reconstruction method considering factors that should be derived preferentially for the geological characteristics of the disposal site and the evaluation of the long-term safety. As a preliminary study on this, we investigated case studies related to geological reconstruction of overseas disposal research institutes, and reviewed which factors are suitable for the domestic granitoid distribution environment. It is expected that systematic and consistent results will be possible in the future through this methodology.

The decommissioning of a nuclear power plant generates large amounts of radioactive waste, which is of several types. Radioactive concrete powder is classified as low-level waste, which can be disposed of in a landfill. However, its safe disposal in a landfill requires that it be immobilized by solidification using cement. Herein, a safety assessment on the disposal of solidified radioactive concrete powder waste in a conceptual landfill site is performed using RESRAD. Furthermore, sensitivity analyses of certain selected input parameters are conducted to investigate their impact on exposure doses. The exposure doses are estimated, and the relative impact of each pathway on them during the disposal of this waste is assessed. The results of this study can be used to obtain information for designing a landfill site for the safe disposal of low-level radioactive waste generated from the decommissioning of a nuclear power plant.