In order to use nuclear energy stably, high level radioactive waste including spent nuclear fuel that is inevitably discharged from nuclear power plants after electricity generation must be managed safely and isolated from the human living area for a long period of time. In consideration of the accumulated amount of spent nuclear fuel anticipated according to the national policy for HLW management, the area required for the deep geological repository facility is expected to be very large. Therefore, it is essential to conduct various studies to optimize the area required for the disposal of spent nuclear fuel in cases where the nationally available land is extremely limited, such as in Korea. In this study, as part of such research, the strategies and the requirements for the preliminary design of a high efficiency repository concept of spent nuclear fuel were established. For PWR spent nuclear fuel, seven assemblies of spent nuclear fuel can be accommodated in a disposal canister, and high burnup of spent nuclear fuel was taken into consideration, and the source terms such as the amount and time of discharge and disposal were based on the 2nd national basic plan. By evaluating the characteristics, the amount of decay heat that can be accommodated in the disposal canister was optimized through the combination of seven assemblies of spent nuclear fuel. The cooling period of the radiation source for the safety assessment of the repository system was set at 55 years, and the operation of the repository would start from 2070 and then the disposal schedule would be conducted according to the disposal scenario based on the national basic plan. With these disposal strategies described above, the main requirements for setting up the conceptual design of the high efficiency repository system to be carried out in this study were described below. • A combination of seven spent nuclear fuels with high heat and spent nuclear fuels with low heat was loaded into a disposal canister, and the thermal limit per disposal canister was 1,600 W. • In order to maintain the long-term performance of the repository, the maximum temperature design limit in the buffer material was set to 130°C. • In the deep disposal environment, the safety factor [yield strength/maximum stress] required to maintain the structural stability of the disposal canister should be maintained at 2.0 or higher so that integrity of the canister can be maintained even under long-term hydrostatic pressure and buffer swelling pressure in the deep disposal environment. • The repository should have a maximum exposure dose of 10 mSv/yr or less, which is the legal limit in case of a single event such as an earthquake, and the risk level considering natural phenomena and human intrusion, which is less than the legal limit of 10-6/yr. These strategies and requirements can be used to develop the high-efficiency geological disposal concept for spent nuclear fuels as an alternative disposal concept.

It is expected that around 576,000 bundles of CANDU spent nuclear fuels (SNF) will be generated from the four CANDU reactors located at the Wolsong site. The authors designed and proposed a reference disposal concept based on the KBS-3 type and KURT geological data in 2022. In addition, we have reviewed the literatures and selected four alternative disposal methods to develop the higherefficiency disposal concept than the reference concept since 2021. As known well, the most important safety functions of the geological disposal are containment and isolation, and the secondary function is retardation. A disposal canister covers the former, and buffer may do the latter. In this study, we design the engineered barrier systems for the four alternative concepts: (1) mined deep borehole matrix, (2) sub-seabed disposal, (3) deep borehole disposal, and (4) multi-level dispoal. Assuming total 10,000 tU of CANDU SNF, four different kinds of unit disposal module consisting of disposal canisters and compacted bentonite buffers are designed based on the technique currently available. Two alternative concepts, sub-seabed disposal and multi-level disposal, share the same unit module design with the reference concept in 2022. For all the alternative concepts, we assume that the density of the compacted buffer is 1.6 g/cm3. For the mined deep borehole matrix disposal, we introduce a disposal canister slightly modified from the Canadian NWMO canister with a capacity of 48 bundles. The thickness of a copper layer is changed to be 10 mm considering the long-term corrosion resistance. The buffer thickness around a disposal canister is 20 cm, and the diameter of a borehole is 100 cm. Two different kinds of buffer blocks are proposed for the easy handling of them. For the deep borehole disposal, a SiC-stainless steel canister is designed, and 63 bundles of CANDU SNF is emplaced in the canister. We expect that the SiC ceramic canister shows very excellent corrosion resistance and has a high thermal conductivity under the geological conditions. The deep borehole will be plugged with four layered sealing materials consisting of granite blocks, compacted bentonite, SiC ceramic, and concrete plugs.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SNFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SNFs with their high radiotoxicity, heat generating, and long-lived radioactivity are currently the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SNF has also been conducted during last couple of decades since around 1997, during which several various type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SNF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide to direct disposal of SNF. The world first Finnish repository will be also this type. Since the characteristics of SNF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SNF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SNF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SNF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development will be roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in near-field, far-field, and even biosphere in and around the conceptual repository system.

As Korea has relatively small land area and large population density compared to other countries considering the DGD concept such as Finland and Sweden, improvements of disposal efficiency in the viewpoint of the disposal area might be needed for the current disposal system to alleviate the difficulties of site selection for the HLW repository. In this research, we conduct a numerical investigation of the disposal efficiency enhancement for a high-level radioactive waste (HLW) repository through three design factors: decay heat optimization, increased thermal limit of buffer, and double-layer concept. In the optimized decay heat model, seven SNFs with the maximum and minimum decay heat depending on actual burn-up and cooling time are iteratively combined in a canister. Thermal limit of buffer is assumed as 100°C and 130°C for reference and high-efficiency repository concepts, respectively. By implementing an optimized decay heat model and a single-layer concept with a thermal limit of buffer set at 100°C, the disposal efficiency increases to 2.3 times of the improved Korean Reference disposal System (KRS+). Additionally, incorporating either an increased thermal limit of buffer to 130°C or a double-layer concept leads to a further 50% improvement in disposal efficiency. By integrating all three design factors, the disposal efficiency can be enhanced up to five times that of the KRS+ repository. Our analysis of rock mass stability reveals that increasing the thermal limit of buffer can generate rock spalling failure in a wider area. However, when accounting for the effect of confining stress by swelling of buffer and backfill using the Mohr-Coulomb failure criteria, the rock mass failure only occurred at the corner between the disposal tunnel and deposition hole when the thermal limit of buffer was increased and a single-layer concept was applied. The results given in this study can provide various options for designing the high-efficiency repository in accordance with the target disposal area and quality of the rock mass in the potential repository site.

The acoustic emission (AE) method as a passive non-destructive monitoring technique is proposed for real-time monitoring of mechanical degradation in underground structures, such as deep geological disposal of high-level nuclear waste (HLW). This study investigates the low-frequency characteristics of AE signals emitted during the fracturing of meter-scale concrete specimens; uniaxial compression tests (UCT) in a lab scale and Goodman jack (GJ) tests in a 1.3 m-long concrete block were conducted while acquiring the AE signals using low-frequency AE sensors. The results indicate a sharp increase in AE energy emission at approximately 60% and 80% of the yield stresses in the UCT and GJ tests, respectively. The collected AE signals were primarily found in two frequency bands: the 4-28 kHz range and the 56-80 kHz range. High-frequency AE signals were captured more as the stress increased in the GJ tests, which was in contrast to the UCT tests. Furthermore, the AE signals obtained from the Goodman jack tests tended to lower RA values than the UCT results. This study presents unique experimental data with low-frequency AE sensors under different loading conditions, which provides insights into field-scale AE monitoring practices.

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.

To obtain a license for a deep geological disposal repository for spent nuclear fuel, it is necessary to perform a safety assessment that quantifies the radiological impact on the environment and humans. One of the key steps in the safety assessment of a deep geological repository is the development of scenarios that describe how the repository evolves over the performance period and how events and processes affect performance. In the field of scenario development, demonstrating comprehensiveness is critical, which describes whether all factors that are expected to have a significant impact on the repository's performance have been considered. Mathematical proof of this is impossible. However, If the scenario development process is logical and systematic, it can support the claim that the scenario is comprehensive. Three primary approaches are being considered for scenario development: ‘Bottomup’, ‘Top-down’, and ‘Hybrid’. Hybrid approach provides a more systematic and structured process by considering both the FEPs (Features, Events, Processes) and safety functions utilized in the bottomup and top-down approaches. Many countries that develop recent scenarios prefer demonstrating scenario comprehensiveness using a hybrid approach. In this study, a systematic and structured scenario development process of a hybrid approach was formulated. Based on this, sub-scenarios were extracted that describe the phenomena occurring in the repository over the performance period, categorized by period. By integrating and screening the extracted sub-scenarios, a scenario describing the phenomena occurring over the entire period of disposal was developed.

As of 2023, there has been significant progress worldwide in the management of nuclear fuel’s spent radioactive waste (HLW). Several countries have made important strides in advancing their plans for the construction of deep geologic repositories (DGRs) to safely dispose of their nuclear waste. Finland led the way, with its nuclear waste management organization, Posiva Oy, submitting an application for an operating license for a DGR for spent fuel generated by the nuclear power plants of its owners. The facility, ONKALO, will be located on the island of Olkiluoto and is expected to begin final disposal in the mid-2020s. Sweden also approved SKB’s application to build a DGR in Forsmark, and an encapsulation plant next to the Clab interim storage facility. In Switzerland, Nagra selected Nordic Lagern as the site for the Swiss DGR, and is preparing the general license applications for the required facilities. Meanwhile, Canada’s Nuclear Waste Management Organization (NWMO) narrowed down the possible locations for its DGR to two, and expects to name its preferred site by fall 2024. The UK established four Community Partnerships to participate in the siting process for a DGR, with Nuclear Waste Services (NWS) responsible for identifying a site. Andra, the French organization responsible for managing all French radioactive waste, is expected to submit an application by the end of the year for a DGR in France that will contain HLW resulting from reprocessing of spent fuel assemblies from French nuclear power plants, as well as intermediate-level waste. Overall, the progress made by these countries represents a tangible and sustainable step forward in the management of spent fuel and HLW, and brings us closer to the safe and effective long-term disposal of nuclear waste.

Various linear system solvers with multi-physics analysis schemes are compared focusing on the near-field region considering thermal-hydraulic-chemical (THC) coupled multi-physics phenomena. APro, developed at KAERI for total system performance assessment (TSPA), performs a finite element analysis with COMSOL, for which the various combinations of linear system solvers and multi-physics analysis schemes should to be compared. The KBS-3 type disposal system proposed by Sweden is set as the target system and the near-field region, which accounts for most of the computational burden is considered. For comparison of numerical analysis methods, the computing time and memory requirement are the main concerns and thus the simulation time is set up to one year. With a single deposition hole problem, PARDISO and GMRESSSOR are selected as representative direct and iterative solvers respectively. The performance of representative linear system solvers is then examined through a problem with an increasing number of deposition holes and the GMRES-SSOR solver with a segregated scheme shows the best performance with respect to the computing time and memory requirement. The results of the comparative analysis are expected to provide a good guideline to choose better numerical analysis methods for TSPA.

As part of the safety case development for generic disposal sites in Korea, it is necessary to develop generic assessment models using various geosphere–biosphere interfaces (GBIs) and potentially exposed groups (PEGs) that reflect the natural environmental characteristics and the lifestyles of people in Korea. In this study, a unique modeling strategy was developed to systematically construct and select Korean generic biosphere assessment models. The strategy includes three process steps (combination, screening, and experts’ scoring) for the biosphere system conditions. First, various conditions, such as climate, topography, GBIs, and PEGs, were combined in the biosphere system. Second, the combined calculation cases were configured into interrelation matrices to screen out some calculation cases that were highly unlikely or less significant in terms of the exposure dose. Finally, the selected calculation cases were prioritized based on expert judgment by scoring the knowledge, probability, and importance. The results of this study can be implemented in the development of biosphere assessment models for Korean generic sites. It is believed that this systematic methodology for selecting the candidate calculation cases can contribute to increasing the confidence of future site-specific biosphere assessment models.

Several countries have been operating radioactive waste disposal (RWD) programs to construct their own repositories and have used natural analogues (NA) studies directly or indirectly to ensure the reliability of the long-term safety of deep geological disposal (DGD) systems. A DGD system in Korea has been under development, and for this purpose a generic NA study is necessary. The Korea Atomic Energy Research Institute has just launched the first national NA R&D program in Korea to identify the role of NA studies and to support the safety case in the RWD program. In this article, we review some cases of NA studies carried out in advanced countries considering crystalline rocks as candidate host rocks for high-level radioactive waste disposal. We examine the differences among these case studies and their roles in reflecting each country’s disposal repository design. The legal basis and roadmap for NA studies in each country are also described. However because the results of this analysis depend upon different environmental conditions, they can be only used as important data for establishing various research strategies to strengthen the NA study environment for domestic disposal system research in Korea.

The major concern in the deep geological disposal of spent nuclear fuels include sulfide-induced corrosion and stress corrosion cracking of copper canisters. Sulfur diffusion into copper canisters may induce copper embrittlement by causing Cu2S particle formation along grain boundaries; these sulfide particles can act as crack initiation sites and eventually cause embrittlement. To prevent the formation of Cu2S along grain boundaries and sulfur-induced copper embrittlement, copper alloys are designed in this study. Alloying elements that can act as chemical anchors to suppress sulfur diffusion and the formation of Cu2S along grain boundaries are investigated based on the understanding of the microscopic mechanism of sulfur diffusion and Cu2S precipitation along grain boundaries. Copper alloy ingots are experimentally manufactured to validate the alloying elements. Microstructural analysis using scanning electron microscopy with energy dispersive spectroscopy demonstrates that Cu2S particles are not formed at grain boundaries but randomly distributed within grains in all the vacuum arc-melted Cu alloys (Cu-Si, Cu-Ag, and Cu-Zr). Further studies will be conducted to evaluate the mechanical and corrosion properties of the developed Cu alloys.

Because of the massive development of nuclear power plants in China in recent years, China is facing the challenge of radioactive waste disposal. China has established complete regulatory requirements for radioactive waste disposal, but it also has encountered problems and challenges in low-level radioactive waste disposal in terms of management, selection of disposal facility sites, and implementation of a site selection plan. Three low-level radioactive waste disposal facilities that have been operated in China are described, and their activity limits, locations, and capacities are also outlined. The connotations of “regional” and “centralized” disposal policies are discussed in light of the characteristics of the radioactive waste. The characteristics and advantages of the regional and centralized disposal policies are compared. It is concluded that the regional disposal policy adopted in 1992 can no longer meet the current disposal needs, and China should adopt a combination of the two disposal policies to solve the problem of radioactive waste disposal.

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y−1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the keff value, 0.95. The STAD canister had the lower keff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in keff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly 3H (500–3,600 Bq·g−1), 14C (7.5–29 Bq·g−1), 55Fe (1.1– 7.1 Bq·g−1), 59Ni (0.60–1.0 Bq·g−1), 60Co (0.74–70 Bq·g−1), 63Ni (0.60–94 Bq·g−1), 90Sr (0.25–5.0 Bq·g−1), 137Cs (0.64–8.7 Bq·g−1), and 152Eu (0.19–2.9) Bq·g−1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g−1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

Concrete waste generated in the result of dismantling a concrete structure in a radiation control area and refractory brick waste generated from uranium pellet sintering furnace are surface-contaminated by uranium particle of which the enrichment is below 5%. These wastes are hard to decontaminate so it was necessary to develop the process for its disposal. So, we developed the Process Control Plan (PCP) for disposal of radioactive concrete waste describing a whole sequence of disposal and inspecting procedures based on the KNF Radioactive Waste Quality Assurance Plan (KN-WQAP) established in 2021. Based on the PCP, we crushed the concrete waste by jaw-crusher. Then we sieved the crushed concrete waste and removed the particle of which size is below 0.3 mm, using sieve-vibrator where the 0.3 mm mesh-sized sieve is installed inside. Before conducting the crush-sieving method based on the PCP, we conducted Process Control Assessment (PCA) based on the KN-WQAP. The purpose of the PCA is to check whether the output of the process satisfies the Acceptance Criteria of Korea Radioactive Waste Agency (KORAD) so that we could confirm the validity of the PCP. The evaluation item of the PCA is a particulate size verification test. The test is passed only if the component ratio of a particle size below 0.2 mm is less than 15% and the particle size below 0.01 mm is less than 1%. The very first 3 drums passed the test, so we began applying the PCP to whole target drums. In the process of conducting the crush-sieving method in earnest, qualified inspectors based on KNWQAP participated conducting sampling, measuring and checking whether a foreign material was included. They tested samples and packaged drums regarding 5 spheres of general, radiological, physical, chemical and biological characteristic. KNF disposed concrete and refractory brick waste by the crush-sieving method so that KNF could take over 100 drums to KORAD in 2021. But, it is needed to be improved that a dust size below 0.3 mm is generated as a secondary waste which needs to be solidified for the final disposal and the work environment is not good enough because of the dust.

The IAEA recommended considerations for exemption regulations of consumer products containing greater amounts of radioactive isotopes than the amounts specified for generic exemption. One of the major considerations is the expected exposure dose should be less than 10 μSv/y and 1 mSv/y for general cases and low probability cases, respectively, in all predictable scenarios. Under this recommendation, many countries evaluated the radiation dose for exposure scenarios of various products in consideration of the national circumstances and, then, established their own specific exemption regulation. In Republic of Korea, the “Regulation on substances excluded from radioactive isotopes” was legislated to specify consumer products excluded from regulation. However, as the usage status and product specifications has changed over time, it is necessary to periodically verify the validity of the regulation criteria in the view of exemption justification. In this study, we developed the use and disposal scenarios in consideration of the domestic use of thorium-containing gas mantle and evaluated radiation dose of each scenario accordingly. The gas mantles are used as a wick for gas lanterns and the maximum activity of natural thorium contained among the currently available gas mantles is 12.5 kBq. Radioactive isotopes in the decay chain of natural thorium can be divided into three groups according to their physical characteristics, and exposure routes suitable for each group were considered in dose calculation. Currently, most gas mantles are installed in camping lanterns. Therefore, we developed use scenarios related to camping. The average number of camping trips and time spent at the campground were set by the data from Korea Tourism Organization. Tent sizes and vehicle specifications were determined by referring to surveys and products in Korea. The used gas mantle is disposed of in a garbage bag for general waste and transported to landfill or incinerator. We determined the amount of gas mantle discarded in landfill and incinerator by the data from Korea Environment Corporation. The exposure time and amount handled by an individual were determined by considering the number of waste collection vehicles, landfills, and incinerators. Although we assumed the maximum activity of the gas mantle for conservative evaluation, the calculated radiation doses for the use and disposal scenarios were below the general requirement (i.e., 10 μSv/y) in all scenarios.

Decommissioning of a nuclear power plant (NPP) generate large amounts of various types of wastes. In accordance with the Nuclear Safety and Security Commission Notice of Korea (No. 2020- 6), they are classified as High Level Waste (HLW), Intermediate Level Waste (ILW), Low Level Waste (LLW), Very Low Level Waste (VLLW) and Exempt Waste (EW) according to specific activities. More than 90% of the wastes are at exempt level, mostly metal and concrete wastes with low radioactivity, of which the concentrations of nuclides is less than the allowable concentration of self-disposal. The self-disposal or recycling of these wastes is widely used worldwide. More than 10,000 drums, based on 200 L drum, are expected to be produced in the decommissioning process of a unit of nuclear power plant. Due to the limited storage capacity of the intermediate & low level waste disposal facility in Gyeongju, recycling and self-disposal of EW are actively recommended in Korea. A variety of scenarios were proposed for recycling and self-disposal of decommissioning metal/ concrete wastes, and a computational program called REDISA was developed to perform the dose evaluation for each recycling and self-disposal scenario. The REDISA computer program can calculate external and internal exposure doses by simulating the exposure pathways from waste generation, thru transport, processing, manufacture, to the final destination of recycling or self-disposal. In this study, the self-disposal scenario was only considered for the dose evaluation. Many studies have been conducted to evaluate the exposure doses of the radioactive waste disposal sites. However, there have been few researches on dose evaluation for self-disposal landfills. In particular, the dose evaluation is important not only during the operation period, but also for a long period after the facility is closed. To this end, we developed a conceptual model for dose evaluation for post-closure scenarios of the self-disposal landfill of decommissioning metal/concrete wastes with reference to the methodology of IAEA-TECDOC-1380. The model incorporates three exposure pathways, including external exposure from contaminated soil, internal exposure by inhalation, and internal exposure by ingestion of water and food grown in contaminated soil. The duration of the dose evaluation is set to 100,000 years after the closure of landfill facility. Co-60 was selected as dominant nuclide, and dose evaluation was performed based on unit specific activity of 1 Bq/g. Exposure doses shall be verified for their application in accordance with the annual dose limit of 10 Sv/yr for self-disposal. As a result, the post-closure scenario of selfdisposal landfills have shown negligible effects on public health, which means that the exposures doses from transportation and operational processes should be considered more carefully for selfdisposal of decommissioning metal/concrete wastes.