Nuclear power plants decommissioning is planned to be started in middle of the 2020. It is necessary to develop safety evaluation and verification technology during decommissioning to ensure the safety of security monitoring measures and maintenance measures, appropriate emergency plans and preparations for decommissioning, and the use of proven engineering when establishing decommissioning plan. For this purpose, a nuclear power plant decommissioning plan is prepared in several stages before decommissioning. When a lifetime of a nuclear power plant has reached, it needs to be decommissioned and therefore operator company should submit decommissioning plans to the National Safety and Security Commission. And safety analysis should be included in this document and it is explained in chapter 6. According to the NSSC Notice No. 2021-10, it is largely divided into principles and standards, exposure scenarios, dose assessment, residual radioactivity, abnormal events, and risk analysis. When unexpected radiological accident is happened, both public and occupational dose analysis should be conducted. However, research on the former can be found easily on the other hands, research on the latter is not active. In this paper, method of choosing scenarios of accidents during the decommissioning the nuclear power plants is briefly introduced. Accidents during nuclear power plants decommissioning cases in USA is chosen and its risk is evaluated by using risk matrix and ranked by AHP method. During the decommissioning phases, varieties of radioactive waste is expected to be generated such as contaminated concrete and metal. On the other hand, Dry Active Waste (DAW) is generated and its amount is and its amount is 7,353 drums. Characteristic of DAW is highly flammable compared to concrete or metal. Moreover, depending on method of radioactive waste conditioning and type of radioactive nuclides, release rate of the nuclides varies. Thus this type of radioactive waste is critical to fire accidents and such accident can occur extra dose exposure which exceeds the guideline of the regulatory body to workers. Therefore, in this paper, occupational dose exposure during the fire accident is conducted.

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.

Radioactive source terms are important factor in design, licensing and operation of SMR (Small Modular Reactor). In this study, regulatory requirements and evaluation methodology for normal operation on NuScale SMR, which received standard design certification approval on September 11, 2020 from US NRC, are reviewed. The radioactive waste management system of nuclear power reactor should be designed to limit radionuclide concentration in effluents and keep radioactive effluents at restricted area boundary ALARA according to 10 CFR 20 and 10 CFR 50 Appendix I. Also, in general, the coolant source term to calculate the off-site radiological consequences for normal operation of SMR should be determined by using models and parameters that are consistent with regulatory guide 1.112, NUREG- 0017 and the guidance provided in ANSI/ANS-18.1-1999, and the result should be corrected by reflecting the design characteristics of SMR. The coolant source term of NuScale, unlike the case of large NPPs, cannot rely solely on empirical source term data, because the NuScale source term is based on first principle physics, operational experience from recent industry, and lessons learned from large PWR operation. Fission products in reactor coolant are conservatively calculated using first principle physics in SCALE Code assuming 60 GWD/MTU. The release of fission products from fuel to primary coolant based on industry operational experience is determined as fuel failure fraction of 0.0066% for normal operation source term and 0.066% for design basis source term while coolant source term of large NPP is calculated by using ANSI/ANS-18.1 for normal operation and fuel failure fraction of 1% for design basis source term. Water activation products in reactor coolant are calculated from first principles physics and corrosion activation products are calculated by utilizing current large PWR operating data (ANSI/ANS 18.1- 1999) and adjusted to NuScale plant parameters. Also, because ANSI/ANS 18.1-1999 is not based on first principle physics models for CRUD generation, buildup, transport, plate-out, or solubility, NuScale has incorporated lessons learned by using ERPI’s primary water chemistry and steam generator guidelines to ensure source term is conservative and design of materials used cobalt reduction philosophy to help ensure the coolant source term are conservative. Based on the coolant source term calculated according to the above-described method, the annual releases of radioactive materials in gaseous and liquid effluents from NuScale reactor are evaluated. Currently, Small Modular Reactors such as ARA, SMART 100 are under review for licensing in Korea. This study will be helpful to understand how the reactor coolant system source terms are defined and evaluated for SMR.

This study presents an example of creating and optimizing a task sequence required in an automated remote dismantling system using a digital manufacturing system. An automated remote dismantling system using a robotic arm has recently been widely studied to improve the efficiency and safety of the dismantling operations. The task sequence must be verified in advance through discrete eventbased process simulation in a digital manufacturing system to avoid problems in actual remote cutting operations as the main input of the automated remote dismantling system. The laser cutting method can precisely cut complicated target structures such as reactor internals with versatility, but a robot and a pre-prepared program are required to deploy sophisticated motion of the laser cutting head on the target structure. For safe and efficient dismantling operations, the robot’s program must be verified in advance in a virtual environment that can represent the actual dismantling site. This study presents creating and optimizing the task sequence of a robotic underwater laser cutting as part of the project of developing an automated remote dismantling system. A task sequence is created to implement the desired cutting path for the target structure using the automated remote dismantling system in the virtual environment. The task sequence is optimized for the posture of the laser cutting head and the robot to avoid collisions during the operation through discrete event-based process simulation since the target structure is complicated and the volume occupied by the laser cutting head and the robot arm is considerably large. The task sequence verified in the digital manufacturing system is demonstrated by experiments cutting the target structure along the desired cutting path without any problems. The various simulation cases presented in this study are expected to contribute not only to the development of the automated remote dismantling system, but also to the establishment of a safe and efficient dismantling process in the nuclear facility decommissioning.

Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

In the pilot scale test, the two scale-up factors (Electric energy per order EEO, Electric energy per mass EEM) were conducted to design the Chemical Waste Decomposition & Treatment System (CWDS). The CWDS consist of two kind UV lamp reactors to improve the decomposition rate of oxalic acid, which are low pressure amalgam UV lamp and medium pressure UV lamp. The two reactors were connected in series, and the hydrogen peroxide is mixed through a line mixer at the front of the reactor and injected into the reactors. The CWDS was connected with the full system decontamination equipment to purify the residual oxalic acid after chemical decontamination process. The full system decontamination equipment were included Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS) to operate the Oxidation/Reduction decontamination process and purify the process water. After decontamination process, the waste water will be cooled down into the 40°C and passed through the UV reactor at 110 gpm with hydrogen peroxide injection. The concentration of waste water is expected oxalic acid 1,700 ~ 2,000 ppm, Iron 5 ~ 20 ppm. As a result of the CBD test in the laboratory with simulated waste liquid, the amount of Low pressure amalgam lamp UV dose required to decompose 95% of oxalic acid in 2 m2 waste water was up to 1,800 mJ/cm2. The amount of medium pressure lamp UV dose was up to 450 mJ/cm2 at the same condition. We conducted demonstration test using 2 m2 waste water after the oxidation/reduction decontamination process, the decomposition rate 95% was obtained by low pressure amalgam UV lamp and medium pressure UV lamp reactor each.

The decommissioning of Kori Unit 1 is expected to generate a large amount of clearance waste. Disposing of a large amount of clearance waste is economically costly, so a recycling method has emerged. However, clearance waste recycling is expected to cause many conflicts among various stakeholders. In the previous study, possible conflicts were selected in consideration of the domestic environment and major issues. Based on this, this study classifies stakeholders involved in conflicts by group, and suggests ways to enhance understanding by stakeholder and enhance reliability. In this study, stakeholders are classified into four groups that share the same conflicts, and each of the following measures is suggested. 1) Stakeholder Engagement. 2) Common understanding of radiation risks, dialogue between the public/recycling industry/ regulatory agency. 3) Incentives to promote recycling clearance waste. 4) Reliable outlet store for recyclable clearance waste. The above understanding enhancement measures are presented so that a solution to conflict can be smoothly derived when designing a clearance waste-related consultative body composed of interested parties in the future. As a more specific solution, measures to enhance stakeholder trust can be suggested for each understanding enhancement measure. Reliability enhancement measures are also presented so that they can be applied to each stakeholder group, and these are as follows. 1) Write a stakeholder engagement plan, Measures for stakeholder participation in measuring the radioactivity concentration of clearance waste. 2) Active use of easy-to-understand radioactivity comparison data, Expansion of information on environmental radiation dose to public, nuclear/radiation education, Held a tour event at the nuclear power plant decommissioning site, New website for clearance waste information disclosure. 3) Incentives for recycling industries in which the Ministry of Environment or KHNP partially bears the losses that occur when the sales rate is low. Incentives are provided to consumers by including recyclables of clearance waste for Green Card’s green consumption points. 4) Online outlets open for recyclable clearance waste with easy-to-understand radioactivity comparison data. It is expected that if the above-mentioned reliability enhancement measures are used, it will be possible to secure the trust of stakeholders and reduce the gap between stakeholders in the future clearance-related consultative body.

The radioactive Sr-90, which is formed from beta decay, is well known as one of the most commonly detected nuclides in radioactive waste. In 2015, it was reported that Sr-90 was observed in some soil and metal wastes among the 516 drums of radioactive waste transferred from the decommissioning site of the Korea Research Reactor (in Seoul) to the disposal site (in Gyeongju). Decontamination and sequestration of radionuclides, including Sr, from nuclear waste is important because they are hazardous and harmful to the ecological environment. Immobilization of these nuclides using a zeolite framework is suitable and simple method that has been widely studied. Therefore, it is still necessary to continuously explore the thermal stability of various zeolites and environmental changes around adsorbed cations in zeolite pore for effective immobilization of these radionuclides. In this study, we observed the thermal stability in fully Sr-exchanged natrolite (Sr-NAT), one of small-pore zeolite, from room temperature to 350°C using the in-situ synchrotron X-ray powder diffraction and thermogravimetric (TGA) analysis. In addition, we investigated the structural changes in Sr-NAT during temperature increase by Rietveld analysis. Sr-NAT exhibited apparent zero thermal expansions (ZTE) with the thermal expansion coefficients of -3(1) × 10-6 at the initial stage of increasing the temperature due to dehydration process. In the section from 250°C to 300°C, a phenomenon like negative thermal expansion (NTE) occurs in which the unit cell volume of Sr-NAT decreases despite the increase in temperature. Sr-NAT maintained well its crystallinity up to 350°C, and it became amorphous at 350°C. In this study, we provide a fundamental understanding of the structural changes and thermal stability mechanism of Sr-exchaged zeolite natrolite with increasing temperature.

An induction melting facility includes several work health and safety risks. To manage the work health and safety risks, care must be taken to identify reasonably foreseeable hazards that could give rise to risks to health and safety, to eliminate risks to health and safety so far as is reasonably practicable. If it is not reasonably practicable to eliminate risks to health and safety, attention have to be given to minimize those risks so far as is reasonably practicable by implementing risk control measures according to the hierarchy of control in regulation, to ensure the control measure is, and is maintained so that it remains, effective, and to review and as necessary revise control measures implemented to maintain, so far as is reasonably practicable, a work environment that is without risks to health or safety. The way to manage the risks associated with induction melting works is to identify hazards and find out what could cause harm from melting works, to assess risks if necessary – understand the nature of the harm that could be caused by the hazard, how serious the harm could be and the likelihood of it happening, to control risks – implement the most effective control measures that are reasonably practicable in the circumstances, and to review control measures to ensure they are working as planned.

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.

3D imaging equipment is essential for automated robotic operations that cut radiologically contaminated structure and transfer segmented pieces in nuclear facility dismantling site. Automated dismantling operations using programmed robotic arms can make conventional nuclear facility dismantling operations much more efficient and safer, so dismantling technologies using robotic arms are being actively researched. Resolving the position uncertainty of the target structure is very important in automated robot work, and in general industries, the problem of position uncertainty is solved through the method of teaching the robot in the field, but at the nuclear facility dismantling site, the teaching method by workers is impossible due to activated target structures. Therefore, 3D imaging equipment is a key technology for a remote dismantling system using automated robotic arms at nuclear facility dismantling site where teaching methods are impossible. 3D imaging equipment available in radioactive and underwater environments is required to be developed for a remote dismantling system using robotic arms because most commercial 3D scanners are available in air and certain 3D scanners available in radioactive and underwater environments cannot satisfy requirements of the remote dismantling system such as measurement range and radiation resistance performance. The 3D imaging equipment in this study is developed based on an industrial 3D scanner available in air for efficient development. To protect the industrial 3D scanner against water and radiation, a housing is designed by using mirrors, windows and shieldings. To correct measurement errors caused by refraction, refraction model for the developed 3D imaging equipment is defined and parameter studies for uncertain variables are performed. The 3D imaging equipment based on the industrial 3D scanner has been successfully developed to satisfy the requirements of the remote dismantling system. The 3D imaging equipment can survive up to a cumulative dose of 1 kGy and can measure a 3D point cloud in the air and in water with an error of less than 1 mm. To achieve the requirements, a proper industrial 3D scanner is selected, a housing and shielding for water and radiation protection is designed, refraction correction are performed. The developed 3D imaging equipment is expected to contribute to the wider application of automated robotic operations in radioactive or underwater environments.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, Offgas system should be properly designed since the fission products in off-gas accelerates the corrosion in reactor structure materials and deteriorates the purity of liquid fuel. The design of off-gas system therefore requires the preliminary study of the behavior of evolved fission products in off-gas units and the development of off-gas model is crucial in developing such system. In this study, we corrected the off-gas illustrative model proposed by ORNL (Nuclear Engineering and Design, vol 385(15) 111529, 2021) by employing physically consistent concept of capture rate of fission product and holdup. For the application of the corrected off-gas model to Chloride-based 6 MW Molten Salt Reactor, major fission products were firstly determined from OpenMC based neutronics calculation and chain reaction related to the major fission products were defined. Based on these data, the holdup behavior of fission products in off-gas units (decay tank, caustic scrubber, Halide trap, H2O trap and charcoal bad) were investigated.

The disposal criteria of the domestic LILW disposal facility specifies that fluidized substances such as the spent resin, the evaporator bottom should be solidified in a physically stable solid form, such as cementation and polymerization. And the solidified form applies requirements for compressive strength, immersion test, thermal circulation test, radiation irradiation test, leaching test, and free standing water measurement test. On the other hand, it is specified that immobilization iss applied to wastes with a total radioactivity concentration of more than 74,000 of radionuclides with a half-life exceeding 20 years among non-homogeneous wastes such as spent filters and DAW, but the test requirements are not applied. Nevertheless, it is necessary for waste generator to establish quality control standards for the manufacture of immobilized solid form through reviewing overseas cases and domestic regulations and technical standards. The test requirements for solidified solid form require measurement of structural stability (compressive strength, immersion, thermal cycling, irradiation test), leachability (leaching test), and free standing water measurement. A characteristic of the immobilized solid form is that it is not mixed with the waste and that the cement medium surrounds the waste. Therefore, the structural soundness is higher than that of the solidified solid mixed with waste. In addition, even when in contact with water, the cement medium blocks the contact between waste and water, thereby preventing the spread of radionuclides. Therefore, considering the characteristics of these immobilized solid form, compressive strength test and free standing water measurement are applied for structural soundness. For other tests, it is determined that application is unnecessary.

In Korea, the NUREG-0017 methodology based on realistic model for reactor coolant concentrations are used to estimate the annual radioactive effluent releases for normal operation of nuclear power plant. The realistic model to estimate the radionuclide concentrations in reactor coolant is formulated as a standard, ANSI/ANS-18.1. This standard has provided a set of the reference radionuclide concentrations and adjustment factors for estimating the radioactivity in the principal fluid systems of target plant. Since ANSI/ANS-18.1 was first published in 1976, it was revised in 1984, 1999, 2016, and most recently in 2020. Therefore, this study analyzed revision history of assessment methodology of radioactive source term of light water reactors, which is ANSI/ANS-18.1. Assessment methodology of radioactive source term given ANSI/ANS-18.1 is by using radionuclide concentrations for reactor coolant and steam generator fluid of the reference plant and adjustment factors, which is modifying radioactive source term according to differences in design parameters between reference plant and target plant. There are three type of reference plant: PWR with u-tube steam generator, PWR with once-through steam generator, and BWR. This study analyzed for PWR with u-tube steam generator. Although the standard was revised, evaluation methodology and formula of adjustment factor have been retained, but some of items have been revised. First revision item is reduction of the number of radionuclides and decrease of radioactive concentration in reactor coolant. In the 1976 version of the standard, there were 71 target radionuclides, but the target nuclides have reduced to 57 in 1984 and 56 after 1999. In the case of radioactive concentration in reactor coolant, as the version of standard was updated, the radioactive concentration of 18 nuclides in 1984, 14 nuclides in 1999, and 25 radionuclides in 2016 was decreased. Most of the radionuclides with decrease radioactivity concentration were fission product, it is resulted from improvement of nuclear fuel performance. Second revision item is change of adjustment factors. After the revision in 2016, the adjustment factors for zinc addition plants using natural or depleted zinc are changed. This study analyzed revision history of evaluation methodology of radioactive source term of light water reactors. Furthermore, result of this study will be contributed to the improvement of understanding of assessment methodology and revision history for the radioactive source term.

The dismantlement of the Kori Unit 1 and Wolsong Unit 1 nuclear power plants is scheduled. Since about 40% of the cost of dismantling nuclear power plants is the cost of disposing of generated wastes, it is important to secure recycling technologies. Among them, low and intermediate level radioactive wastes are made of porous filters and adsorbent materials of ceramic foam to remove nuclides such as C-14, I, and Xe generated during nuclear dismantling. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. Therefore, in this study, the foam filter performance was improved by applying a sacrificial material to increase the specific surface area and porosity of the ceramic foam filter. The sacrificial material is burned out with polyurethane (PU) of the green filter before the heat treatment temperature to increase the strength of the ceramic foam filter so that it can be maintained as pores, thereby improving the specific surface area and porosity. The sacrificial materials and melting temperature (Tm) reviewed in this study were anthracite (530~660°C), PMMA (160°C), Cellulose acetate (260~270°C), and aluminum particle (660°C), and their effect on the manufacture of foam filters was studied by applying this. The specific surface part and porosity of the foam filter were improved when anthracite and aluminum particle were added, and PMMA and Cellulose acetate, which are relatively low temperature melting points, were burned out at a temperature lower than PU, and thus their physical properties were not greatly affected. The physical properties and specific surface part and porosity of ceramic foam filters manufactured using various sacrificial materials will be discussed.

Organic complexing agents may affect the mobility of radionuclides at low- and intermediate-level radioactive waste repositories. Especially, isosaccharinic acid (ISA) is the main cellulose degradation product under high pH conditions in cement pore water. ISA can combine with radionuclides and form stable complexes that adversely influence adsorption in the concrete phase, resulting in radionuclides to leach to the near- and far-fields of repositories. This study focuses on investigating the sorption of ISA onto engineered barriers such as concrete, thereby studying adsorption isotherms of ISA on concrete and comparing various isotherm models with the experimental data. The adsorption experiment was conducted in three background solutions, groundwater (adjusted to pH 13 using NaOH), State 1 (artificial cement pore water, pH 13.3), and State 2 (artificial cement pore water, pH 12.5), in a batch system at a temperature of 20°C. Concrete was characterized using BET, Zeta-potential analyzer, XRD, XRF, and SEM-EDS. ISA concentrations were detected using HPLC. The experimental data were best fitted to one-site Langmuir isotherm; On the other hand, either two-site isotherm or Freundlich isotherm couldn’t give reasonable fitting to the experimental data. The observed ISA sorption behavior on concrete is crucial for the disposal of radioactive waste because it can significantly lower the concentration of ISA in the pore water. Although one-site Langmuir isotherm might effectively represent the sorption behavior of ISA on concrete, the underlying mechanism is still unknown, and further investigation should be done in the near future.

Encapsulation using cement as a solidification method for disposal of radioactive waste is most commonly used in most advanced countries in the nuclear technology to date due to its advantages such as low material cost and accumulated technology. However, in case of cement solidification, since moisture or hydroxyl group in cement is decomposed by radioactivity, it may happen that gas is generated, structural stability is weakened, and leachability is increased due to low chemical durability. Therefore, the various new solidification methods are being developed to replace it. As one of these alternative technologies, for dispersible metal compounds generated through the incineration replacement process, the study on engineering element technology for powder metallurgy is under way, which overcomes the interference problem between mechanical elements and media that may occur during the process such as the homogeneous mixing process of the target powder substance and additives used in the powder metallurgy concept-based sintering process for the solidification of the final glass composite material (GCM), the process of creating a compressed molded body using a specific mold, the process of final sintering treatment. The solidification process of dispersible radioactive waste can be largely divided into pre-treatment stage, molding stage, and sintering stage, and the characteristics of the final radioactive waste solidification material can vary depending on the solidification treatment characteristics of each stage. In relation with these characteristics, the matters to be considered when designing device for each stage to solidify dispersible radioactive waste (property of super-mixing device for homogenized powder formation, structural geometry and pressure condition of molding device for production of compressed molded body, temperature and operation characteristics of sintering device for final glass composite material (GCM), etc.) are drawn out. It is expected that the solidification device design reflecting these considerations will meet all disposal conditions of radioactive waste material, such as compressive strength and leaching characteristics of solidified radioactive waste material, and create a uniformized solidification of radioactive waste material.

In this study, the process of compressing/packaging the spent filters of Kori Unit 1, which was conceptually presented in the previous study, is advanced so that disposal suitability for each step can be secure efficiently. In particular, the differences between the previous study and this study are that the disposable filters are screened using an In-Situ Object Counting System (ISOCS), and the method of collecting representative samples for development of scaling factor is specified. The process of compressing/packaging the spent filters consists of 7 stages as follows. 1) Collecting: The spent filters temporarily stored in the filter room are collected by dose and type remotely using a robot system to minimize the radiation exposure of workers according to a pre-established packaging plan. 2) Screening: The gamma activity concentration of the spent filters received by the robot system is measured by ISOCS. The spent filters below the low-level waste concentration limit and the surface dose are transferred into the compression system, while the others are returned in the filter room again. 3) Sampling: The external perforator drilling/cutting the filter was developed for sampling required for the new scaling factors. Since the sampling is collected remotely, the risk of exposure to workers can be reduced. The newly developed scaling factor will be used to verify the disposal suitability of the packages. 4) Compression: According to the pre-established plan, the spent filter collected by dose and type, is supplied to the compression system considering the dose and radionuclide inventory. Whether to additionally store the compressed filter in the drum is determined by checking the accumulated dose. 5) Immobilization: Immobilization with a safety material is necessary when inhomogeneous wastes, like spent filters, have the total radionuclide concentration with a half-life of more than 20 years is 74,000 Bq/g or more and for filling rate or non-dispersible treatment of particulates. 6) Packaging and Analysis: Waste information is labelled onto the package after the measurements of surface dose rate and surface contamination. Finally, using the drum assay system, the gamma radionuclide concentration is measured to identify at least 95% of the total radioactivity concentration of the package. 7) Temporary Storage and Delivery: The packages are moved to temporary storage in the plant prior to disposal. After establishing the plan for delivery and applying for a takeover request to KORAD, if the acceptance inspection is passed, the packages are transported to the disposal facility.