In this research, the dose rate was measured using a backpack-type scan survey device at 4 sites in sites around Nuclear Power Plants (Kori, Wolsong, Hanbit, Hanul), and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. Kori, Wolsong and Hanul power plants were measured within 2 km of the power plant, and Hanbit power plants were measured about 6.7 km from the power plant. As a result of measuring the dose rate with a backpacktype scan survey device, the average dose rate was the lowest in the measurement site 1 at 0.090 μSv/h, and the highest in the measurement site 4 at 0.145 μSv/h. All measurement points showed the domestic environmental dose rate level. The data obtained by the scan survey was visualized using the classed post and gridding functions of the surfer program. As a result of measurement with the HPGe detector, 137Cs was not detected, and only natural nuclides were detected. Among the detected natural nuclides, the radioactivity ratio was the highest for 40K with an average of 94.56%, and the lowest for 214Pb with an average of 0.26%. The results of this research can be used as basic data for radiation environment surveys around nuclear power plants. Further studies are needed to evaluate the radiation impacts by region and environment through periodic measurements.

The US NRC developed a program called NRCDose3 to evaluates the environmental impact of radiation around nuclear facilities. The NRCDose3 code is a software suite that integrates the functionality of three individual LADTAP II, GASPAR II, and XOQDOQ Fortran codes that were developed by the NRC in the 1980’s and have been in use by the nuclear industry and the NRC staff for assessments of liquid effluent and gaseous effluent, and meteorological transport and dispersion, respectively. Through the integrated program, it is possible to conduct safety assessment and environmental impact assessment from liquid and gaseous effluent when operating permits are granted. In addition to a more user-friendly graphic user interface (GUI) for inputting data, significant changes have been made to the data management and operation to support expanded capabilities. The basic calculation methods of the LADTAP II, GASPAR II, and XOQDOQ have not been changed with this update to the NRCDose3 code. Several features have been added. The previous program used only ICRP-2 dose conversion factor, but the new program can additionally use dose conversion factor of ICRP-30 and ICRP-72. In the previous program, 4 age groups (infant, child, teen, and adult) were evaluated during dose evaluation, but when ICRP-72 was selected, 6 age groups (infant, 1-year, 5-year, 10-year, 15-year, and adult) could be evaluated. In addition, when selecting ICRP-72, many user-modifiable parameters such as food intake and exposure time were added. It will be referred to E-DOSE60, a program currently under development.

After the Fukushima nuclear power plant accident in 2011, interest in technology for evaluating residents’ exposure to effluents generated from nuclear power plants at the time of the accident has increased. KHNP has developed the S-REDAP program and is using it to evaluate radiation dose and recommend resident protection measures in the event of a nuclear power plant emergency. Its main functions are source term evaluation, atmospheric diffusion evaluation, radiation dose evaluation, etc. Based on these evaluations, resident protection measures are evaluated. In Japan, evaluation is conducted through a program called SPEEDI-MP (System for Prediction of Environmental Emergency Dose Information Multi-model Package) created by JAEA (Japan Atomic Energy Agency). Similar to S-REDAP, the program also evaluates effluents emitted from nuclear facilities through source term evaluation and atmospheric diffusion factor evaluation. In JAEA, through a program using SPEEDI-MP, the source term evaluation was performed in collaboration with NSC (Nuclear Safety Commission) in the event of the Fukushima nuclear plant accident, and dose evaluation in Japan was performed 2 months as an atmospheric diffusion factor using meteorological data for 2 days. Through comparative analysis of evaluation data from Japan, improvements to the current program be derived.

K-DOSE60, a off-site dose calculation program currently used by khnp, is performing evaluation based on the gaseous effluent evaluation methodology of NRC Reg. Guide 1.109. In particular, H-3 and C-14, which are the major nuclides of gaseous effluent, are evaluated using a ratio activity model. Among them, H-3 is additionally evaluating the dose to OBT (Organically Bound Tritium) and HT as well as HTO (Triated water). However, NRC Reg. Guide 1.109 is a methodology developed in the 1970s, and verification was performed by applying the evaluation methodology of H-3 and C-13 presented by IAEA TRS-472 in 2010 to the current K-DOSE60. The IAEA TRS-472 methodology also includes OBT and HT for H-3. In order to apply the ratio radioactivity model presented in IAEA TRS-472, the absolute and relative humidity were calculated using the weather tower of the nuclear site and used for H-3 evaluation. For the dose evaluation of HT, the previously used Canada Chalk River Lab. (CNL) conversion factor was used. For atmospheric carbon concentration, the carbon concentration presented in IAEA TRS-472 was used, not the carbon concentration in the 1970s of NRC Reg. Guide 1.109. It was confirmed that the K-DOSE60, which applied the changed input data and methodology, was satisfied by performing comparative verification with the numerical calculation value.

In 2022 and 2023, the Korea Institute of Nuclear Safety (KINS), a regulatory body, revised the regulatory guidelines for off-site dose evaluation to residents, marine characteristics surveys around nuclear facilities, and environmental radiation surveys and evaluation around nuclear facilities. In addition, the NRC, a US regulatory body, has revised regulatory guide 1.21 (MEASURING, EVALUATING, AND REPORTING RADIOACTIVE MATERIAL IN LIQUID AND GASEOUS EFFLUENTS AND SOLID WASTE) to change environmental programs for nuclear facilities. The domestic regulatory guidelines were revised and added to reflect the experience of site dose evaluation for multiple units during the operation license review of nuclear facilities, the resident exposure dose age group was modified to conform to ICRP-72, and the environmental monitoring plan was clarified. In the case of the US, the recommended guidelines for updating the long-term average atmospheric diffusion factor and deposition factor, the clarification of the I-131 environmental monitoring guidelines for drinking water, and the clarification of the procedures described in the technical guidelines when changing environmental programs have been revised and added. Through such regulatory trend review, it is necessary to preemptively respond to changes in the regulatory environment in the future.

Radioactive materials depositied after nuclear accident or radiological emergency result in radiation exposure to individuals living in long-term contaminated territories. Therefore, the remedial actions should be taken on affected areas for the evacuated residents to return to their homes and normal lifestyle. Meanwhile, radiation exposure occurs through various pathways by work types during the site clean-up. Therefore, dose assessment is crucial to protect emergency workers and helpers from the potential radiological risk. This study estimated the exposure dose to individuals decontaminating the areas contaminated with 60Co, 63Ni, 90Sr, 134Cs, 137Cs, and then calculated the maximum workable soil concentration to comply with the reference level of 20 mSv/y for transition to existing exposure situations. For the realistic assessment, the detailed exposure scenarios depending on the types of work (excavation, collection, transportation, disposal, landfill), and the relevant exposure pathways were used. In addition, with the LHS (Latin Hypercube Sampling) - PRCC (Partial Rank Correlation Coefficient) method, sensitivity analysis was performed to identify the influence of the input parameters and their variation on the model outcomes. As a result, the most severe exposure-induced type was identified as the excavator operation with an annual individual dose of 4.75E-01 mSv at the unit soil concentration (1 Bq/g), from which the derived maximum workable soil concentration was 4.21E+01 Bq/g. Dose contribution by isotopes were found to be 60Co (55.63%), 134Cs (32.01%), and 137Cs (12.28%), and the impact of 63Ni and 90Sr were found to be negligible. Dose contribution by exposure pathways decreased in the following order: ground-shine, soil ingestion, dust inhalation, and skin contamination. Furthermore, the most high sensitive input parameters and their PRCC were found to be as the dilution factor (0.75) and as the exposure time (0.63). In conclusion, the results are expected to contribute to optimize radiation protection strategeis for recovery workers and to establish appropriate response procedures to be applicable in areas with high deposition density after a radiological or nuclear emergency.

For licensees who face the decommissioning project for the first time, even if they can utilize their experience in operation, they should be well prepared and assessed for the risks of dismantling activities reflecting the characteristics of decommissioning. This can be included in the risk management of the decommissioning project, but what we want to discuss in this study is the evaluation of the industrial risk of the actual work before the dismantling work is carried out. We would like to focus more on the review of dismantling activities subject to industrial risk assessment and a series of processes for risk assessment. The dismantling work plan will need to obtain approval from the supervisory department before work on the Systems, Structures, and Components (SSCs) can be carried out. At this time, risk assessment may be included among many safety-related required documents, which are divided into radiological and non-radiological risks. The target activities at Level 1 level can include preparation for dismantling and maintenance of facilities, dismantling big components, removing the contamination of concrete structures, managing radioactive waste, etc. In addition, it can be composed of preparation work, removal of connections, lifting/installation, cutting, radiation/radioactivity measurement, and withdrawal as detailed work stages of each item’s activities. For domestic nuclear decommissioning projects, two major performance organizations, licensees and contractors, must be considered. Regarding risk assessment, the licensee will have a supervisory department controlling decommissioning activities and an HSE department at the site, and a process will need to be established in consideration of the contractor’s work organization. Therefore, activities in the risk assessment process may be established. In this study, risk assessment was reviewed as safety-related matters to be considered when carrying out the dismantling work. Safety-related risk assessment is a necessary procedure for performing practical dismantling activities, and this should be considered well in advance. Therefore, work activities and criteria were established for risk assessment, and the performance process was assumed to apply them. In terms of the performance organization and the responsibilities and roles of the processes to be performed by each organization were constructed, and this can be referred to in the process of preparing for the decommissioning project.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.

Kori Unit 1 was permanently shut down in 2017 and is preparing to be dismantled. Decommissioning nuclear power plants is expected to generate a lot of decommissioning waste. Therefore, a radioactive waste treatment complex will be built on the site to safely and effectively the process of decommissioning waste generated from the Kori Unit 1, and the details are specified in the decommissioning plan. Therefore, a safety assessment should be conducted according to the facility’s normal and abnormal operations to construct a radioactive waste treatment complex. Currently, a safety assessment for a radioactive waste treatment complex can be conducted by the Safety Assessment Framework (SAFRAN) Tool based on the Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) methodology developed by the International Atomic Energy Agency (IAEA). The SAFRAN Tool can be calculated radiation dose and hazard quotient (HQ) for workers and the public under normal and abnormal conditions of the radioactive waste treatment complex. When evaluating the radiation dose for the public due to releasing radioactive materials into the air or discharging radioactive materials into liquids, the radiation dose is calculated using the amount discharged or released from the treatment complex, and the Pathway Dose Factors (PDFs) derived from the generic environmental model given in the IAEA Safety Reports Series No.19. PDFs, which reflect the specific site data rather than the generic environmental model data, should be calculated and evaluated when performing the safety evaluation of the radioactive waste treatment complex to be built on the Kori site. In addition, in the SAFRAN tool, there is an inconvenience in that it must be calculated separately by radionuclides to calculate the contribution of dose or HQ for each radionuclide. Therefore, in this study, a safety assessment tool for a radioactive waste treatment complex was developed using Visual Basic by supplementing the limitations of the SAFRAN tool. This tool was developed to allow users to choose whether to apply PDFs based on the IAEA SRS-19 based on the generic environmental model or PDFs calculated to reflect the specific site data. Furthermore, the tool considered all types of decommissioning wastes that may occur during the decommissioning of the Kori Unit 1 and the treatment process scheduled to be introduced. Therefore, this study is expected to be used as basic data when conducting the safety assessment of radioactive waste treatment complex scheduled to be introduced in Korea.

Concerns with colloids, dispersed 1~1,000 nm particles, in the LILW repository are being raised due to their potential to enhance radionuclide release. Due to their large surface areas, radionuclides may sorb onto mobile colloids, and drift along with the colloidal transport, instead of being sorbed onto immobile surfaces. To prevent adverse implications on the safety of the repository, the colloidal impact must be evaluated. In this paper, colloid analysis done by SKB is studied, and factors to be considered for the safety assessment of colloids are analyzed. First, the colloid generation mechanism should be analyzed. In a cementitious repository, due to a highly alkaline environment, colloid formation from wastes may be promoted by the decomposition of organic materials, dissolution of inorganic materials, and corrosion of metals. Radiolysis is excluded when radionuclide inventory is moderate, as in the case of SKB. Second, colloid stability should be evaluated to determine whether colloids remain in dispersion. Stable colloids acquire electric charges, allowing particles to continuously repel one another to prevent coagulation. Thus, stability depends on the pH and ionic condition of the surroundings, and colloid composition. For instance, under a highly alkaline cementitious environment, colloids tend to be negatively charged, repelling each other, but Ca2+ ion from cement, acting as a coagulant, makes colloid unstable, promoting sedimentation. As in the case of SKB, the colloidal impact is assumed negligible in the silo, BMA, and BTF due to their extensive cement contents, but for BLA, with relatively less cement source, the colloidal impact is a potential concern. Third, colloid mobility should be assessed to appraise radionuclide release via colloid transport. The mobility depends on the density and size of colloids, and flow velocity to commence motion. As a part of the assessment, the filtration effect should also be included, which depends on pore size and structure. As in the case of SKB, due to static hydraulic conditions and engineering barriers, acting as efficient filters, colloidal transport is expected to be unlikely. In the domestic underground repository, the highly alkaline environment would lead to colloid formation, but due to high Ca2+ concentration and low flow velocity, colloids would achieve low stability and mobility, thus colloidal impact would be a minor concern. In the future, with further detailed analysis of each factor, waste composition, and disposal condition, reliable data for safety evaluation could be generated to be used as fundamental data for planning waste acceptance criteria.

The engineered barrier system (EBS) for deep geological disposal of high-level radioactive waste requires a buffer material that can prevent groundwater infiltration, protect the canister, dissipate decay heat effectively, and delay the transport of radioactive materials. To meet those stringent performance criteria, the buffer material is prepared as a compacted block with high-density using various press methods. However, crack and degradation induced by stress relaxation and moisture changes in the compacted bentonite blocks, which are manufactured according to the geometry of the disposal hole, can critically affect the performance of the buffer. Therefore, it is imperative to develop an adequate method for quality assessment of the compacted buffer block. Recently, several non-destructive testing methods, including elastic wave measurement technology, have been attempted to evaluate the quality and aging of various construction materials. In this study, we have evaluated the compressive wave velocity of compacted bentonite blocks via the ultrasonic velocity method (UVM) and free-free resonant column method (FFRC), and analyzed the relationship among compressive wave velocity, dry density, thermal conductivity, and strength parameter. We prepared compacted bentonite block specimens using the cold isostatic pressure (CIP) method under different water content and CIP pressure conditions. Based on multiple regression analysis, we suggest a prediction model for dry density in terms of manufacturing conditions. Additionally, we propose an empirical model to predict thermal conductivity and unconfined compressive strength based on compressive wave velocity. The database and suggested models in this study can contribute to the development of quality assessment and prediction techniques for compacted buffer blocks used in the construction of a disposal repository.

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).

A variety of microorganisms are contained in the groundwater and surrounding environment at the depth of a deep geological repository, and could adversely affect the integrity and/or safety of the facility under certain thermal, hydraulic and chemical conditions. In particular, microbial activity (in the buffer and backfill) around the canister can cause corrosion of the canister through sulfide production by sulfate-reducing bacteria (SRB), and subsequently promote radionuclide release through the corroded part. Namely, this phenomenon is important in a perspective of performance assessment since it will have an impact on the post-closure exposure dose in the biosphere by accelerating radionuclide leakage into the near-field due to deterioration of the canister integrity In Finland, the performance assessment on microbial activity in buffer, backfill, and plug was performed for the licensing. However, in Korea, researches relevant to microbial activity are only in the early stage as of now. Accordingly, in this study, we draw initial considerations for the performance assessment on the phenomenon in the domestic facility based on review results for the methodology carried out as part of operating license application (i.e. SC-OLA). Studies on the performance assessment of microbial activity in Finland were mainly performed: (a) to investigate complex interactions among microorganisms in the repository by analyzing both indigenous and exogenous microorganisms through drilling, geological and geochemical analysis, (b) to identify microbial interactions at the buffer, backfill, and host rock interface for specific microorganisms that may affect activity of other microorganisms and integrity of the repository, (c) to analyze canister corrosion caused by microbial activity, particularly sulfide production by SRB, and (d) to characterize microbial illitization of montmorillonite that could affect permeability, hydraulic conductivity, and structural integrity of the repository. From reviewing studies above, it is judged that studies labelled as (b) through (d) are applicable to the performance assessment of microbial activity for the domestic facility regardless of specific conditions in Korea. However, for study labelled as (a), the following data on reflecting domestic conditions should be additionally obtained: (1) radionuclide inventory and temperature in spent nuclear fuel, (2) swelling pressure and organic carbon content of bentonite, and (3) size, shape, and gas composition of pores in bentonite. Results of this study could be directly applied to the design and performance assessment for buffer and backfill components, provided that input data specific to the domestic disposal facility is prepared for the assessment required.

Bentonite, a material mainly used in buffer and backfill of the engineering barrier system (EBS) that makes up the deep geological repository, is a porous material, thus porewater could be contained in it. The porewater components will be changed through ‘water exchange’ with groundwater as time passes after emplacement of subsystems containing bentonite in the repository. ‘Water exchange’ is a phenomenon in which porewater and groundwater components are exchanged in the process of groundwater inflow into bentonite, which affects swelling property and radionuclide sorption of bentonite. Therefore, it is necessary to assess conformity with the performance target and safety function for bentonite. Accordingly, we reviewed how to handle the ‘water exchange’ phenomenon in the performance assessment conducted as part of the operating license application for the deep geological repository in Finland, and suggested studies and/or data required for the performance assessment of the domestic disposal facility on the basis of the results. In the previous assessment in Finland, after dividing the disposal site into a number of areas, reference and bounding groundwaters were defined considering various parameters by depth and climate change (i.e. phase). Subsequently, after defining reference and bounding porewaters in consideration of water exchange with porewater for each groundwater type, the swelling and radionuclides sorption of bentonite were assessed through analyzing components of the reference porewater. From the Finnish case, it is confirmed that the following are important from the perspective of water exchange: (a) definition of reference porewater, and (b) variations in cation concentration and cation exchange capacity (CEC) in porewater. For applying items above to the domestic disposal facility, the site-specific parameters should be reflected for the following: structure of the bedrock, groundwater composition, and initial components of bentonite selected. In addition, studies on the following should be required for identifying properties of the domestic disposal site: (1) variations in groundwater composition by subsurface depth, (2) variations in groundwater properties by time frame, and (3) investigation on the bedrock structure, and (4) survey on initial composition of porewater in selected bentonite The results of this study are presumed to be directly applied to the design and performance assessment for buffer and backfill materials, which are important components that make up the domestic disposal facility, given the site-specific data.

In buffer, a main component of engineering barrier system (EBS) in the deep geological repository, mass loss is mainly caused by upheave and mechanical erosion. The former is a phenomenon that bentonite in the upper part of the buffer moves to the backfill region due to groundwater intake and swelling. And, the latter is a phenomenon that bentonite on the surface of the buffer moves to the backfill region due to groundwater flow at the interface with host rock as the buffer saturates. Buffer mass loss adversely affects the fulfilment of the safety function of the buffer that is to limit and retard radionuclide release in the event of canister failure. Accordingly, in this paper, we reviewed how to consider this phenomenon in the performance assessment for the operating license application in Finland, and tentatively summarized data required to conduct the analysis for the domestic facility based on the review results. Regarding buffer mass loss, the previous studies carried out in Finland are categorized as follows: 1) experiment on the amount of buffer upheave with groundwater inflow rate (before backfilling), 2) analysis for the amount of buffer upheave with groundwater inflow rate (after backfilling), 3) analysis of buffer erosion rate with groundwater inflow rate, 4) analysis for distribution of the groundwater inflow rate into the buffer for all deposition holes (using ConnectFlow modeling results), and 5) analysis of buffer mass loss with groundwater salinity. Finally, the buffer mass loss distribution table was derived from the results of 1) through 3) by combining with that of 4). Given these studies, the following will be required for the performance assessment for buffer mass loss in the domestic disposal facility: a) distribution table of buffer mass loss for combined interactions taking into account effect of 5) (i.e. 1), 2), 3), and 5) + 4)), and b) Threshold for buffer mass loss starting to negatively affect the fulfilment of the safety function of the buffer. Even though it is judged that the results of this study could be directly applied to developing the design concept of EBS and to conducting the performance assessment in the domestic disposal facility, it is essential to prepare a set of input data reflecting the site-specific design features (e.g. dimension, material used, site, etc.), which include saturation time and groundwater salinity.

In the deep geological repository, a considerable quantity of cementitious materials is generally used for structural stability of subcomponents such as grout and concrete plug of disposition tunnel. Strong alkaline leachates (pH>13) are produced after cement is dissolved by groundwater inflow from bedrock. When the leachates are transported to bentonite porewater (e.g. buffer and backfill) and thereby water exchange occurs, the physical properties of bentonite such as swelling capacity and hydraulic conductivity are changed, which eventually affects the safety function and long-term stability of engineered barrier system (EBS). Thus, in this paper, we reviewed the performance assessment methodology for cement-bentonite interaction in the operating license application for the Finnish deep geological repository, and suggested what to prepare for the analysis on the domestic disposal facility. In Finland, thermal-hydraulic-chemical analysis for dissolution of montmorillonite by alkaline leachates resulting from cement degradation during the saturation of bentonite was carried out using PRECIP code. From this analysis, it was confirmed that effect on pH was considered to be more significant than that on temperature and bentonite saturation. As a result of this analysis, it was predicted that all primary minerals (including montmorillonite, quartz, and calcite) were dissolved and some secondary minerals (e.g. chalcedony and celadonite) was precipitated by alkaline cement leachates transported to the bentonite. In addition, it was shown that silica was preferentially released while the montmorillonite was dissolved, thus cementation of the bentonite was occurred. Through this phenomenon, the swelling capacity of bentonite is reduced and the hydraulic conductivity of bentonite is increased, which have a significant impact on the performance of the buffer and backfill. Considering this, study on spreading of alkaline leachates, which is a condition for dissolution of montmorillonite, is necessary for the performance assessment of the domestic deep geological repository. However, this requires the site-specific data for the following in the disposal site: (a) distribution in fractured bedrock and pore structure (e.g. porosity, pore size distribution and pore morphology) in the bedrock, (b) hydraulic gradient and salinity concentration of groundwater, and (c) flux and velocity of groundwater. Results of this study is considered to be directly utilized to the conceptual design and performance assessment of the deep geological repository in Korea, provided that additional data on microbiological properties of groundwater are obtained for the site selected.