As the decommissioning of Kori Unit 1 progresses, securing technology for treatment and disposal of radioactive wastes that have not been disposed of so far, such as spent filters, is recognized as an urgent task. In this study, a method of confirming the disposal suitability of spent filters was presented by reviewing the waste characteristics as presented in the waste acceptance criteria (WAC). The waste characteristics to be satisfied to ensure disposal suitability of waste are largely classified into general requirements, solidification and immobilization requirements, radiological requirements, physical requirements, chemical requirements, and biological requirements. First, the general requirement is to prove that the prohibited waste form has not been introduced into items related to waste form and packaging, and to confirm the suitability of disposal through step-by-step packaging photos, generation information, X-ray inspection, and visual inspection. Second, in the solidification and immobilization requirements, spent filters are non-homogeneous waste, and if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1 or more, they must be immobilized. Third, in order to meet the characteristic criteria for nuclides and radioactivity concentration, sampling and scaling factors development are required and based on this, nuclides must be identified and demonstrated to be below the disposal concentration limits. Surface dose rate and surface contamination should be measured in accordance with standardized procedures and disposal suitability should be confirmed through document tests recording the measured values. Fourth, in order to satisfy the physical requirements of the particulate matter and filling rate characteristics, the spent filter must be immobilized, if necessary, thereby ensuring disposal suitability. Meanwhile, free water in the spent filter should be removed through pre-drying and dehydration, and the disposal suitability should be confirmed by applying a test. Fifth, the criteria for chelating agents should be checked for disposal suitability through operation records and component analysis of spent filters, and documents, that can prove harmful substances are removed in advance and no harmful substances are included in the package, should be provided. Lastly, in biological requirements, if the spent filters contain corrosive or infectious substances, they should be removed in advance and disposal suitability should be confirmed by providing documents that can prove that such substances are not included in the package.

This study established a process to ensure the disposal suitability of spent filters stored in the untreated state in Kori unit 1 and presented the following procedures and requirements for confirming the disposal suitability for each process. The process for securing spent filter disposal suitability consists of collecting spent filters, compression, immobilization, analysis and packaging, and storage stages. The requirements for confirming the acceptance criteria for each process are as follows. (1) Collecting: Since the high radioactivity spent filters are being stored in the filter room of Kori unit 1, those are collected by a remote system to minimize the exposure dose of workers due to spent filter handling. In order to satisfy the surface dose rate requirements, spent filters with a surface dose rate of 10 mSv·hr−1 or more are classified and collected, and stored temporary storage place until a separate treatment plan is determined. The checkpoints in this process are the surface dose rate, etc. (2) Compression: The collected spent filters are analyzed gamma nuclides such as Co-60 and Cs-137, using a field-applicable nuclide analyzer, and then applying the scaling factors to determine whether it is disposable. Spent filters whose radioactivity concentration is confirmed to be less than the disposal concentration limit is compressed into compression ratios determined by surface dose rate. The checkpoints in this process are nuclide information, surface dose rate, compression ratio, spent filter loading quantity, etc. (3) Immobilization: A spent filter is a non-homogeneous waste that is immobilized with a proven safety material such as cement if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1. Meanwhile, immobilization of inhomogeneous waste can be considered to satisfy disposal criteria such as particulate matter and filling rate. The checkpoints in this process are the immobilizing material, filling rate, etc. (4) Analysis and Packaging: Immobilized drums shall be determined to be 95% or more of the total radioactivity of waste packages by measuring the radioactivity concentration of nuclides using a nuclide analysis device. Finally, measure the surface dose rate and surface contamination of the package, and attach the package label recording the identification number, date, total radioactivity, surface dose rate, and surface contamination information to the packaging container. (5) Storage: Packaging containers are moved to and stored in a temporary waste storage or storage area before disposal.

Radioactive waste generated during the decommissioning of Kori Unit 1 can be packaged in a transport container under development and transported to a disposal facility by sea transport or land transport. In this study, the cost of each transport method was evaluated by considering the methods of land transport, sea transport, and parallel transport of the radioactive waste dismantled at Kori Unit 1. In evaluating the shipping cost, the shipping cost was evaluated by assuming the construction of a new ship without considering shipping by CHEONG JEONG NURI, which is currently carrying operational waste. Since the cargo hold of CHEONG JEONG NURI was built to fit the existing operating waste transport container and is not suitable for transporting the transport container currently under development, sea transport using CHEONG JEONG NURI was excluded in this paper. In the case of on-road transportation, the final fare for each distance was calculated in accordance with the Enforcement Decree of the Freight Vehicle Transportation Business Act, and the cost of onroad transportation was evaluated by estimating the labor cost of the input manpower required for onroad transportation. The cost of on-road transportation was estimated to be approximately KRW 510 million, the product of the total number of transports 459 times the sum of the cost of transportation vehicle freight cost of about KRW 720,000 and the labor cost of input personnel of KRW 380,000. It is difficult to predict the cost of building a new ship at this point, as the cost of building new ship is determined by the cost of number of items such as ship design specifications and material prices, labor costs, and finance costs at the time of construction. Accordingly, considering the 2% annual inflation rate based on the shipbuilding cost (about KRW 26 billion) and financing cost (about KRW 12 billion) at the time of construction of the CHEONG JEONG NURI (2005 yr.), decommissioning of Kori Unit 1 (2025 yr.) construction cost finance cost was estimated and evaluated. According to the result of comparing the transport cost for each transport scenario, land transport is about 510 million won, which is advantageous in terms of economic feasibility compared to the sea transport scenario. However, when transporting by land, it is disadvantageous in terms of acceptability of residents because it is transported multiple times on general roads. The cost of building a new ship is about KRW 56.4 billion, which is disadvantageous in terms of the cost of transporting waste from the dismantling of Kori Unit 1. But, in the future, cost reduction can be expected if waste materials issued when dismantling nuclear power plants are transported.

Currently, in domestic nuclear power plants (NPP), the spent filters (SFs) used for the purpose of reducing and purifying the radiation of the primary cooling water system are temporarily stored in an untreated state. In order to dispose of SFs, radioactive nuclide analysis (RNA) of SFs is required to be conducted. As segmented gamma scanner (SGS) is already being used in Kori NPP, utilizing SGS for RNA of SFs would be practical and economical. In this paper, factors required to be considered to improve accuracy of SGSs for RNA of SFs are studied. The analysis of the nuclide inventory of the packaging drum for radioactive waste should be performed by the indirect drum nuclide analysis method. The material of the SFs is iron (SS304) on the outside, and paper on the inside. In addition, to meet disposal acceptance criteria, radioactive waste drums are packaged in thick grouting or shielding drums. Therefore, it is necessary to derive an appropriate correction method for high inhomogeneity and thick media. Considering these factors, evaluating radionuclides inventory plans to measure gamma rays in SGS mode. Correct the gamma ray measurement by examining the medium attenuation factor and error factors. In this way, the inventory of gamma nuclides is calculated, and the specific radioactivity of beta ray and alpha particle emitting nuclides other than gamma rays is planned to be calculated by applying scaling factors.

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.

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. Thus, evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, safety assessments conducted in Sweden and UK are studied, and the factors required to be considered for appropriate safety assessment of cellulose is analyzed. SKB (Sweden) conducted safety assessment of cellulose degradation as a part of long-term safety assessment of SFR. SKB determined that ISA would impact sorption of trivalent and tetravalent radionuclides (Eu, Am, Th, Np, Pa, Pu, U, Tc, Zr and Nb) at concentration higher than 10−4–10−3 M, and impact sorption of divalent radionuclides (Ni, Co, Fe, Be and Pb) at concentration higher than 10−2 M. Then, SKB conservatively set the upper limit of ISA concentration to be 10−4 M and conducted cellulose degradation evaluation on each waste package type, considering the expected disposal environment of SFR. Based on the calculated results, some of the waste packages showed concentration of ISA to be higher than 10−4 M, so SKB conservatively developed waste acceptance criteria to prevent ISA being produced to an extent of affecting the safety of the repository. SKB conducted safety assessment only for the repositories with pH above 12.5 and excluded 1BLA from the safety assessment as the expected pH of 1BLA is around 12, which is insufficient for cellulose to degrade. However, SKB set disposal limit for 1BLA as well, to minimize potential impact in future. Serco (UK) conducted safety assessment of cellulose degradation for the conceptual repository, which is a concrete vault with cementitious backfill. Serco estimated that the pH of repository would maintain around 12.4. Serco conservatively assumed that the pH would be sufficient for cellulose degradation to occur partially, and suggested application of appropriate degradation ratio for safety assessment of cellulose degradation. To conduct appropriate safety assessment of cellulose degradation, an appropriate ISA concentration limit based on radionuclide inventory list, and an appropriate cellulose degradation ratio based on the pH of disposal environment should be determined. As for guidance, below pH 12.5, cellulose degradation is not expected, and between pH 12.5–13, partial cellulose degradation is expected. In future, this study could be used as fundamental data to evaluate safety of the repository.

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.

When the decommissioning of a nuclear power plant begins in earnest, starting with Kori Unit 1, it is necessary to dispose of intermediate-level wastes such as high-dose waste filters and waste resin stored in the power plant, as well as the internal structures of the reactor. However, there are no intermediate-level waste disposal facilities in Korea, and the maintenance of acceptance criteria considering the physical, chemical, and radiological characteristics of intermediate-level waste is insufficient. In this paper, in preparation for the establishment of domestic intermediate-level waste treatment/disposal and acceptance standards, the following major foreign countries’ legal and institutional standards for intermediate-level waste are reviewed, and based on this, factors to be considered when establishing domestic intermediate-level waste treatment/disposal standards were derived. First, although the USA does not define and manage intermediate-level wastes separately, low-level wastes were separated into Class A, B, and C, where land disposal is allowed, and GTCC, which does not allow land disposal. However, it was recently confirmed that the position was changed to recognize the possibility of land disposal of GTCC waste under the condition that the dose to inadvertent intruders does not exceed 5 mSv·yr−1 and a barrier against inadvertent intrusion valid for 500 years is installed. Second, Sweden classifies intermediate-level wastes into short-lived and longlived intermediate-level wastes. The maximum dose rate permitted on packages are different for each vault and a silo of the SFR where short-lived wastes; 100 mSv·h−1 or less is disposed of in BMA, 10 mSV·h−1 or less in BTF, 2 mSv·h−1 or less in BLA and 500 mSv·h−1 or less in silo. Meanwhile, a repository for long-lived low and intermediate level waste, SFL, which could contains significant amounts of nuclides with a half-life greater than 31 years, operations are planned to commence in 2045. Third, France also manages short-lived intermediate-level wastes and long-lived intermediatelevel wastes separately, and the short-lived intermediate-level wastes were disposed of together with short-lived low-level wastes at the La Manche and L’Aube repository. France announced the Cigéo Project, a high- and medium-level long-lived waste plan in 2012, and submitted the creation authorization application for in 2021 with the goal of operating a repository in 2025. Finally, the UK defines intermediate-level waste as “waste whose activity level exceeds the upper limit for low-level waste but does not require heating, which is considered in the design of storage or disposal facilities” and established NIREX to provide deep disposal of intermediate-level radioactive waste. In Finland, wastes with radioactive concentrations of 1 MBq/kg to 10 GBq·kg−1 are classified as intermediatelevel wastes, and a repository was constructed and operated in a bedrock of about 110 m underground. Because the domestic classification standard simply classifies intermediate-level waste as waste exceeding the activity level of low-level waste limit, not high-level wastes, it is necessary to establish treatment and disposal standards by subdividing them by dose rate and long-lived radionuclides concentration to safely and efficiently dispose of intermediate-level waste for. Additionally, there is a need to decide whether or not to reflect safety by inadvertent intruders when evaluating the safety of intermediate-level disposal.

Glass fiber, which was used as an insulation material in pipes near the steam generator system of nuclear power plants, is brittle and the size of crushed particles is small, so glass fiber radioactive waste (GFRW) can cause exposure of workers through skin and breathing during transport and handling accidents. In this study, Q-system which developed IAEA (International Atomic Energy Agency) for setting the limit of radioactivity in the package is used to confirm the risk of exposure due to an accident when transporting and handling GFRW. Also, the evaluated exposure dose was compared with the domestic legal effective dose limit to confirm safety. Q-system is an evaluation method that can derive doses according to exposure pathway (EP) and radioactivity. Exposure doses are calculated by dividing into five EP: QA, QB, QC, QD, and QE. Since the Q-system is used to set the limit of radioactivity that the dose limits is satisfied to nearby workers even in package handling accidents, the following conservative assumptions were applied to each EP. QA, QB are external EP of assuming complete loss of package shielding by accident and radiation are received for 30 minutes at 1 m, QC is an internal EP that considers the fraction of nuclides released into the air and breathing rate during accident, and QD is an external EP that skin contamination for 5 hours. Finally, QE is an internal and external EP by inert gases (He, Ne, Ar, Kr, Xe, Rn) among the released gaseous nuclides, but the QE pathway was excluded from the evaluation because the corresponding nuclide was not present in the GFRW products used for evaluation. In this study, the safety evaluation of GFRW was performed package shielding loss and radioactive material leakage due to single package accident according to assumption of four pathways, and the nuclide information used the average radioactivity for each nuclide of GFRW. As a result of the dose evaluation, QA was evaluated as 2.73×10−5 mSv, QB as 1.06×10−6 mSv, QC as 7.53×10−3 mSv, and QD as 2.10×10−6 mSv, respectively, and the total exposure dose was only 7.56×10−3 mSv, it was confirmed that when compared to the legal limits of the general public (1 mSv) and workers (20 mSv) 0.756% and 0.038%, respectively. In this study, it was confirmed that the legal limitations of the general public and workers were satisfied evens in the event of an accident as a result of evaluating the exposure dose of nearby targets for package shielding loss and radioactive material leakage while transporting GFRW. In the future, the types of accidents will be subdivided into falling, fire, and transportation, and detailed evaluation will be conducted by applying the resulting accident assumptions to the EP.

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.

It has been discovered that the isosaccharinic acid (ISA) formed in a cellulose degradation leachate were capable of forming soluble complexes with thorium, uranium (IV) and plutonium. Since 1993, the ISA has received particular attention in the literature due to its ability to complex a range of radionuclides, potentially affecting the migration of radionuclides. ISA is formed as a result of interactions between cellulosic materials within the waste inventory and the alkalinity resulting from the use of cementitious materials in the construction of the repository. In an alkaline cementitious environment, cellulose degrades mainly via a peeling-off reaction. The main degradation product is ISA, a polyhydroxy type of ligand forming stable complexes with tri- and tetravalent radionuclides. ISA can have an adverse effect on the sorption of radionuclides to an extent which depends on its concentration in the cement pore water and potentially enhance their mobility. The concentration of ISA is governed by several factors such as cellulose loading, cement porosity, extent of cellulose degradation, etc. The sorption of ISA on cement, however, is the process which governs the concentration of ISA in the pore water. According to the experimental result from a literature, the ISA concentration in facilities with a cellulose loading of 5% is calculated to be of the order of 10−4 M. At this level, the effect of cellulose degradation products on radionuclide sorption is negligibly small. Recently in Korea, cellulous limits as waste acceptance criteria is studying and planning to prepare the detailed requirement for near surface radioactive waste disposal facilities. It is desirable to suggest consideration on cellulose disposal limits around the time that the regulatory body and concern organizations establish the cellulose disposal limits as follows. Firstly, identify the cellulose effect on the sorption of the nuclides as cementitious disposal environments such as affected nuclides, threshold value and contribution to radiological risks under domestic disposal environment. Secondly, make sure and consider the difference between lab-scale experimental conditions and probability occurring in real disposal conditions such as probability for generation and persistence of pH in cellulosic material disposal conditions and cellulosic material disposal methods. Finally, consider characterization of cellulosic material such as polymerization, contents of cellulose in law material and time of degradation process. As a result, desirable cellulose limits are to set up for both safety and economic aspect.

Deep geological disposal (DGD) of spent nuclear fuels (SNF) at 500 m–1 km depth has been the mainly researched as SNF disposal method, but with the recent drilling technology development, interest in deep borehole disposal (DBD) at 5 km depth is increasing. In DBD, up to 40SNF canisters are disposed of in a borehole with a diameter of about 50 cm, and SNF is disposed of at a depth of 2–5 km underground. DBD has the advantage of minimizing the disposal area and safely isolating highlevel waste from the ecosystem. Recently, due to an increasing necessity of developing an efficient alternative disposal system compared to DGD domestically, technological development for DBD has begun. In this paper, the research status of canister operation technology and plans for DBD demonstration tests, which subjects are being studied in the project of developing a safety-enhancing high-efficiency disposal system, are introduced. The canister operation technology for DBD can be divided into connection device development and operation technology. The developing connection device, emplacing and retrieving canisters in borehole, adopted the concept of a wedge thus making replacement equipment at the surface unnecessary. The new connection device has the advantage of being well applied with emplacement facilities only by simple mechanical operation. The technology of operating a connection device in DBD can be divided into drill pipe, coiled tubing, free-drop, and wireline. The drill pipe is a proven method in the oil industry, but requiring huge surface equipment. The coiled tubing method uses a flexible tube and shares disadvantages as the drill pipe. The free-drop is a convenient method of dropping canister into a borehole, but has a weakness in irretrievability in an accident. Finally, the wireline method can be operational on a small scale using hydraulic cranes, but the number of operated canisters at once is limited. The test facility through which the connection device is to be tested consists of dummy canister, borehole, lifting part, monitoring part, and connecting device. The canister weight is determined according to the emplacement operation unit. The lifting part will be composed following wireline consisting of a crane, a wire and a winding system. The monitoring part will consist of an external monitoring system for hoists and trolleys, and an internal monitoring system for the connection device’s location, pressure, and speed. In this project, a demonstration test will be conducted in a borehole with 1km depth, 10 cm diameter provided by KAERI to verify operation in the actual drilling environment after design improvement of the connecting device. If a problem is found through the demonstration test, the problem will be improved, and an improved connection device will be tested to an extended borehole with a 2 km depth, 40 cm diameter.