애긴노린재는 긴노린재과에 속하며 한국을 포함한 동아시아 국가의 다양한 곡물 및 관상용 식물의 주요 해충으로 여겨진다. 본 연구에서는 애긴노린재의 17,367 bp 미토콘드리아 유전체에서 13개의 protein-coding genes, 22개의 transfer RNA genes, 2개의 ribosomal RNA genes 과 non-coding A+T rich region를 확인하였다. G+C content는 23%로 나타났고 다른 긴노린재과와의 염기서열 유사성이 N. cymoides (94.5%), N. fuscovittatus (91.7%)으로 높은 것을 발견하였다. 애긴노린재의 미토콘드리아 유전체 정보는 향후 긴노린재과의 진화 연구와 해충 방제를 위한 정보로 널리 사용될 수 있다.

This study was performed to evaluate the separation of Sr, Cs, Ba, La, Ce, and Nd using gas pressurized extraction chromatography (GPEC) with anion exchange resin for the quantitation of Neodymium. GPEC is a micro-scaled column chromatography system that provides a constant flow rate by utilizing nitrogen gas. It is overcome the disadvantages of conventional column chromatography by reducing the volume of elution solvent and shortening the analysis time. Here, we compared the conventional column chromatography and the GPEC method. The whole analysis time was decreased by nine times and radioactive wastes were reduced by five times using the GPEC system. Anion exchange resin 1-X4 (200~400 mesh size) was used. The sample was prepared at a 0.8 M nitric acid in methanol solution. The elution solvent was used at a 0.01 M nitric acid in methanol solution. Finally the eluate was analyzed by ICP-MS to determine the identification and recovery. In this case, we applied the natural isotopes of LREEs (139La, 140Ce, and 144Nd) and high activity nuclides (88Sr, 133Cs, and 138Ba) instead of radioactive isotopes for the preliminary test; as a result, unnecessary radioactive waste was not produced. The recoveries were 93.9%, 105.9%, 91.9%, 47.6%, 35.9%, and 79.9% of Sr, Cs, Ba, La, Ce, and Nd, respectively. The reproducibility of recoveries by GPEC were in the range 2.8%–10.9%.

Safety-related items in the decommissioning Nuclear Power Plants (NPPs) can largely consider safety for workers and residents. At this time, the effects of radioactive contamination on the Systems, Structures, and Components (SSCs) are caused by the performance of work related to Decontamination and Dismantlement (D&D) activities. Classification according to dismantling activities will be important, and the decay factor of radionuclides and the impact of contaminations due to plant characteristic (thermal and electrical capacity) in estimation of exposure dose from such activities will be considered compared to other overseas NPPs. Therefore, this study will consider some factors to consider for comparison with overseas cases in estimating worker exposure dose. To assess worker exposure doses, the classification of decommissioning activities must first be made. It should be classified including large components that can be generally considered, and the contents should be similar to compare with overseas cases. In case of decommissioned NPPs with prior experience, it is possible to predict worker’s exposure with respect to plant capacity, but this does not seem to have a specific correlation when reviewing the related data. Depending on the plant capacity, the occurrence of contamination of radioactive materials may have some correlation, but it cannot be determined that it has causality with the worker’s dose when dismantling. In addition, it is expected that the effects of workers’ exposure doses will vary depending on when the highly contaminated SSCs will be dismantled from permanent shut down. Therefore, the decay correlation coefficient for this high radiation dose works should be considered. If the high radiation dose work is performed before the base year, a correlation coefficient larger than 1 value will be applied, and in the opposite case, a value less than 1 will be applied. Whether or not to perform Full System Decontamination (FSD) is also an important consideration that affects worker dose, and correlation factors should be applied. In this study, the matters to be considered when estimating worker dose for dismantling NPPs were reviewed. This suggests factors to be reflected in the work classification and dose results for comparison with overseas NPP experiences. Therefore, when doing the workers’ dose estimation, it is necessary to derive a normalized doses considering each correlation factor when comparing with overseas cases along with dose estimation for the dismantling activities.

Detectors used for nuclear material safeguards activities are using scintillator detectors to quickly calculate the uranium enrichment at various nuclear material handling facilities. In order to measure the uranium enrichment, a region of interest is set around 185.7 keV which is the main gamma emission energy of uranium-235 in which the proportional relationship between the amount of uranium-235 and the net count is used. It is necessary to perform channel/energy calibration that a specific channel of the multi-channel analyzer is set to 185.7 keV. Most detector manufacturers have a built-in calibration source so that it is automatically performed when the detector starts to operate. In addition, the scintillator detector requires attention because the channel/energy gain may change depending on the ambient temperature so that a calibration source is used to compensate for this. In this paper, the spectral features are examined from among the scintillator detectors seeded with calibration sources used for safeguards activities. For this purpose, FLIR’s Identifinder-2 R400 T2 model and Canberra’s NAID model were used. HM-5 contains about 15nCi of Cs-137 and a photoelectric peak occurs at 662.1 keV. NAID contains about Am-241 of 55 nCi which alpha decays and subsequently emits gamma rays of 59.5 keV and 26.3 keV. The major difference among the detectors occurs in the background spectrum due to the difference in the source. From that kind of spectral features, it can be confirmed that the equipment is operating properly only when the spectrum by the corresponding calibration source is accurately known. The results of this study will enable a better understanding of the characteristics of scintillator detectors used for uranium enrichment analysis. Therefore, it is expected to be used as basic research for related software utilization as well as development in the future.

Since 2018, Central Research Institute of Korea Hydro & Nuclear Power (KHNP–CRI) has been operating an X-ray irradiation system with a maximum voltage of 160 kV and 320 kV X-ray tube to test personal dosimeters in accordance with ANSI N13.11-2009 “Personnel Dosimetry Performance- Criteria for Testing”. This standard requires that dosimeters for the photon category testing be irradiated with the X-ray beams appropriate to the ISO beam quality requirements. KHNP-CRI has implemented the fourteen X-ray reference radiation beams in compliance with ISO-4037-1, 2, and 3. When installing the X-ray irradiation system, KHNP-CRI evaluated the uncertainties of dose conversion coefficients for deep and shallow doses, based on “Catalogue of X-ray spectra and their characteristic data – ISO and DIN radiation qualities, therapy and diagnostic radiation qualities, unfiltered X-ray spectra” published by Physikalisch Technische Bundesanstalt (PTB). A CdTe detector (X-123, AMPTEK) with disk type collimators made of tungsten was used to acquire X-ray spectra. The detector was located at 1 m from the center of the target material in the Xray tubes. Six uncertainty factors for the dose conversion coefficients for the fourteen X-ray beams were chosen as follows; the minimum and maximum cut-off energies Emin and Emax, the air density (ρ), the accuracy of the high-voltage of the X-ray tube, statistics of the pulse height spectra and the unfolding method. For example, uncertainty of each quantity for a HK30 beam was calculated to be 0.3%, 2.32%, 0.19%, 1.25%, and 0.13%, and 0.18%, respectively. The combined standard uncertainty for the deep dose conversion coefficient of the HK30 beam was calculated to be 2.67%. The coverage factor corresponding to a 95 percent confidence interval was obtained as k = 1.8 using a Monte Carlo method, which is slightly lower the coverage factor of k = 1.95 for a Gaussian distribution. This seems to result from that two dominant uncertainties, the unfolding uncertainty and minimum cut-off energy uncertainty, follow a rectangular distribution.

At Nuclear Power Plant (NPP), aging management is performed as part of the Periodic Safety Review (PSR) in accordance with the Nuclear Safety Act. The purpose of the aging management program (AMP) is to manage the integrity of structures, systems and components (SSCs) in NPPs over time and use. Through this, aging deterioration is mitigated to increase equipment life and secure long-term operation safety. Fuel Oil Chemistry is one of the AMPs. Through this program, aging management is performed for storage tanks, piping and other metal components that contact with diesel fuel oil. The program is focused on managing loss of material due to general, pitting, crevice, and microbiologically-influenced corrosion (MIC) and fouling that leads to corrosion of the diesel fuel tank internal surfaces. The fuel oil aging management method currently applied to NPPs in Korea measures the concentration of water and particulate contamination in the oil, analyzed the trend, and periodically cleans and inspect the inside of tanks. Among them, in monitoring MIC, a direct analysis and monitoring of the amount of microorganisms may be more effective. In this study, a method for improving the MIC monitoring system for diesel fuel oil systems was reviewed by reviewing reference documents including NUREG 1801 and examining the methods actually applied in US NPPs.

Around the world, Nuclear Power Plants (NPPs) have been operated since the 1950s and are used as a major power source. In Korea, Kori unit 1 stared commercial operation for the first time in 1978, and as of 2023, 25 units of NPPs are in operation. NPPs produce electricity for about 40 to 60 years after receiving an operating license, and after securing safety through a safety evaluation, the operating period is extended. NPPs that operate for a long time are systematically evaluated for safety at regular intervals through Periodic Safety Review (PSR) recommended by the IAEA. In Korea, PSR has been introduced and performed since 2000. This study reviewed the process of the PSR by comparing with the international PSR procedure. The PSR process is established through the IAEA SSG-25 document and proceeds in the order of establishment of basis document - individual factor evaluation - global assessment - integrated improvement plan. In Korea, PSR is carried out in a similar process, but there are some differences from the IAEA’s procedure. The safety factor review is conducted under the agreement of basis document between the licensee and the regulatory body, but the prior agreement procedure with the regulatory body is not reflected in Korea. As a result, if the licensee and the regulatory body have different opinions on the current licensing basis and the modern safety standards after the evaluation is performed, a difference may occur in the review results and safety enhancement items, which may lead to inefficient PSR progress. PSR is conducted for the continuous safe operation and management of NPPs, and it is important to refer to overseas standards and cases. Although procedures, guidelines, and regulatory requirements are in place in Korea, continuous review and improvement are required. It is necessary to improve procedures such as basis document and global assessment in order to more efficiently carry out PSR evaluation by regulatory agency and licensee’s safety enhancement actions of domestic NPPs

Among domestic Nuclear Power Plants (NPPs), there are a total of 10 nuclear power plants whose operating license expires by 2030, excluding Kori unit 1 and Wolsong unit 1, which are permanently shut downed. Continued operation of these nuclear power plants is being reviewed as a government task. For continued operation, nuclear power plant owners must prepare periodic safety review and other evaluation reports to receive reviews to maintain safety even during continued operation. In the safety evaluation of NPP, it is important to refer to overseas cases and operation experiences. In this study, the matters of radioactive waste management for continued operation of NPP was considered by analyzing the safety evaluation reports and safety enhancements of license renewal of NPP in USA, Radioactive waste generated from NPPs can be classified into solid, liquid, and gaseous states. Radioactive waste generated during the operation and maintenance of power plants is classified, stored and treated in the radioactive waste management system according to the source. Equipment and monitors related to radioactive waste management are continuously operated, managed, inspected according to standards and maintain their original functions. Various activities to reduce the generation and emission of radioactive waste from NPPs are performed. After reviewing the NRC’s safety evaluation report on the application documents for license renewal of US NPPs (Sequoyah, Byron and braidwood) the evaluation details and matters requiring enhancement for the radioactive waste management system were confirmed. As a major check, selective leaching occurred in the body of the gray cast iron valve and the heat exchanger shell containing the copper alloy exposed to the radioactive waste liquid. Selective leaching causes loss of material and may interfere with the original function of the facility, so management is required. For the safe operation and management of NPPs, it is important to refer to overseas cases and experiences. Among the safety evaluations for the continued operation of domestic NPPs, in the field of the radioactive waste management system, if the case of the US NPP is referred to, the review by the regulatory body and the action taken by the licensee will be more efficient.

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

In this study, four technologies were selected to treat river water, lake water, and groundwater that may be contaminated by tritium contaminated water and tritium outflow from nuclear power plants, performance evaluation was performed with a lab-scale device, and then a pilot-scale hybrid removal facility was designed. In the case of hybrid removal facilities, it consists of a pretreatment unit, a main treatment unit, and a post-treatment unit. After removing some ionic, particulate pollutants and tritium from the pretreatment unit consisting of UF, RO, EDI, and CDI, pure water (2 μS/cm) tritium contaminated water is sent to the main treatment process. In this treatment process, which is operated by combining four single process technologies using an inorganic adsorbent, a zeolite membrane, an electrochemical module and aluminumsupported ion exchange resin, the concentration of tritium can be reduced. At this time, the tritium treatment efficiency of this treatment process can be increased by improving the operation order of four single processes and the performance of inorganic adsorbents, zeolite membrane, electrochemical modules, and aluminum- supported ion exchange resins used in a single process. Therefore, in this study, as part of a study to increase the processing efficiency of the main treatment facility, the tritium removal efficiency according to the type of inorganic adsorbent was compared, and considerations were considered when operating the complex process.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

The nuclear power plant decommissioning project inevitably considers time, cost, safety, document, etc. as major management areas according to the PMBOK technique. Among them, document management, like all projects, will be an area that must be systematically managed for the purpose of information delivery and record maintenance. In Korea, where there is no experience in the decommissioning project yet, data management is systematically managed and maintained during construction and operation. However, if the decommissioning project is to be launched soon, it is necessary to prepare in consideration of the system in operation, what difference will occur from it in terms of data management, and how it should be managed. As a document that can occur in the decommissioning project, this study was considered from the perspective of the licensee. Therefore, the types of documents that can be considered at Level 1 can be divided into (1) corresponding documents, (2) project documents, (3) internal documents, and (4) reference materials. Four document types are recommended based on Level 1 for the classification of documents to be managed in the decommissioning of nuclear facilities. In this study, documents to be managed in the decommissioning project of nuclear facilities were reviewed and the type was to be derived. Although it was preliminary, it was largely classified into major categories 1, middle categories 2, and 3 levels, and documents that could occur in each field were proposed. As a result, it could be largely classified into corresponding documents, project documents, internal documents, and reference materials, and subsequent classifications could be derived. Documents that may occur in the decommissioning project must be managed by distinguishing between types to reduce the time for duplication or search, and the capacity of the storage can be efficiently managed. Therefore, it is hoped that the document types considered in this study will be used as reference materials for the decommissioning project and develop into a more systematic structure.

The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

The operation and decommissioning of nuclear power plants (NPPs) creates waste in the process of handling radioactively contaminated material, which must be disposed of in a repository or can be recovered of in the same way as conventional waste in the course of handling radioactively contaminated materials. For buildings or sites of NPPs it also has to be decided under what conditions they can continue to be used for other, conventional purposes or demolished. This decision is referred to as “release from supervision under nuclear and radiation protection law” or “clearance” in short. The clearance levels applicable in Germany according to the Radiation Protection Ordinance have been defined such that a radiation dose (hereinafter referred to as “dose”) of 10 μSv per year is not exceeded. The vast majority of the materials resulting from the dismantling of a nuclear power plant (e.g. most of the massive concrete structures) are neither contaminated nor activated. Thus, there is no need to treat these materials as radioactive waste. Emplacement of uncontaminated masses which in Germany is essentially several million tonnes of building rubble in a repository would require additional construction of such facilities, which, in view of the negligible hazard potential, from the point of view of the Nuclear Waste Management Commission (ESK) is clearly to be rejected both economically and, in particular, ecologically. Alternative ways are increasingly discussed in public, such as the abandonment of buildings after gutting, i.e. refraining from demolition of the controlled area buildings of NPPs. Also, another proposal discussed in public, the landfilling or the long-term storage of cleared material at the site, does not offer any safety-related advantages either in the view of the ESK. If, after completion of all dismantling work, the building has been decontaminated such that the clearance levels for buildings are complied with further use of the building rubble resulting from demolition is harmless from a radiological point of view. For these reasons, Germany has deliberately decided to use clearance as an essential measure in the dismantling of NPPs. If it is intended to conventionally reuse or depose of virtually contaminant-free material from controlled areas, it must first undergo a clearance procedure. The prerequisites that must be fulfilled for clearance are regulated in the Radiation Protection Ordinance, which includes two basic clearance pathways: unrestricted and specific clearance. In the following, the basic process of clearance is briefly presented and illustrated for a better understanding. It comprises five steps. Step 1-Radiological characterization by sampling, Step 2-Dismantling of plant components in the controlled area, Step 3- Decontamination, Step 4-Decission measurements, Step 5-Clearacnce and further management. The entire clearance process is governed by a clearance notice and is carried out under the supervision of the competent authority under nuclear and radiation protection law or the independent authorized expert commissioned by it. The clearance pathways contained in the Radiation Protection Ordinance have proven themselves in practice. They permit safe and proper management of material from dismantling and release of the site from supervision under nuclear and radiation protection law. These German regulatory procedures should be taken into account and deregulation and removal should be used as appropriate and necessary tools in the process of decommissioning NPPs in order to return non-hazardous materials to the material cycle or for conventional disposal.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

The spent filters used to purify radioactive materials and remove impurities from primary systems at nuclear power plants (NPPs) have been stored for long periods in filter storage rooms at NPPs due to concerns about the unproven safety of the treatment method, absence of disposal facilities, and risk of high radiation exposure. In the storage room at Kori Unit 1, there are approximately 227 spent filters of 9 different types. The radiation dose rates of filters range from 0.01 to 500 mSv/hr. Recently, a comprehensive plan has been established for the treatment and disposal of radioactive waste that has not yet been treated to facilitate decommissioning of NPPs. As a follow-up measure, compression and packaging optimization processes are being developed to treat the spent filters. KHNP plans to dispose of the spent filters after compressing, packaging, and immobilizing them. However, the spent filters are currently stored without being sorted by type or radiation intensity. If the removal and packing of the filters are done randomly without a plan for the order of withdrawal and subsequent processes, issues may arise such as a decrease in drum loading efficiency and exceeding the dose limit of the package. In this study, the number of drums needed to pack the spent filters was calculated, considering the filter size, weight, quantity, dose rate, shielding thickness of drum, and loadable quantity in a shielding drum (SD). Then, the spent filters that can be loaded on each drum were classified into one group. In addition, the withdrawal order for each group was set so that the filter withdrawal, compression, and packaging processes could be performed efficiently. The spent filter groups are as follows: (1) compression/12 cm SD (17 groups), (2) compression/16 cm SD (6 groups), (3) non-compression/ intermediate storage container (17 groups, additional radiation attenuation required due to high dose rate), and (4) unclassified (5 groups, determined after measurement due to lack of filter information). The withdrawal order of the groups was determined based on several factors, including visual identification of the filter, ease of distribution after withdrawal, work convenience, and safety. Due to the decay of radioactivity over time, the current dose rate of the spent filters is expected to be much lower than at the time of waste generation. Therefore, in the future, sample filters will be taken from the storage room to measure their radioactivity and radiation dose rate. Based on these measurements, a database of radiological characteristics for the 227 filters will be created and used to revise the filter grouping.

It is important to make a strategy for clearance-level radioactive waste. Sampling and disposal plans should be drawn up with characteristics of target waste. In this paper, a target clearance-level radioactive waste is used in a laboratory for experiments with Cs-137 and Co-60, unsealed radioactive sources with gamma radiation isotopes. Therefore, it is enough to analyze with HPGe to check the contaminant level. The laboratory fume hood combined multiple materials, which means some are volume contamination and others are surface contamination. The wood, plastic, and drywall boards, which are absorbent volume contaminated parts and make up PVC pipes, base cabinet doors, backside baffles, etc., will be sampled with coring methods. The metals and glasses, which are unabsorbent, surface-contaminated parts, are sampled with smear methods. The work surface, baffles, exhaust plenum, and glass sash inside parts have a high possibility of being contaminated. The hood body, flame, base cabinet, PVC pipe (the rare end of the filter), and blower transition case have a low possibility of becoming contaminated. When we checked with HPGe, except for the work surface (which was below clearance level), other parts were less than MDA. The highest radionuclide concentration was in PVC pipe: Cs-137C 3.91E-02 (Bq/g), Co-60 4.54E- 03 (Bq/g). It is less than clearance level. Therefore, the waste was applied for the clearance level radioactive wastes and got permission from the regulatory body.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.