By developing plasma torch melting technology in 1996, our company has developed the first generation 150 kW (’96~’02), the second generation 500 kW (’08~’12), and the third generation MW plasma torch melting facility (’14~’18), and completed facility upgrading (’20~’23). The MW plasma torch melting facility is equipped with CCTV to monitor waste input, melting, torch integrity, and melt discharge. The lens is installed inside a metal housing made of stainless steel to prevent damage caused by external impacts and high temperatures, and supplies nitrogen to prevent cooling and lens contamination. As a result of the demonstration test, as the temperature inside the melting furnace increased after starting the plasma torch, the resolution decreased along with noise in the CCTV, and facility monitoring was difficult due to high temperatures and foreign substances (fume). Based on the test results, CCTV was changed to a non-insertion type that was not directly exposed to high temperatures, and a filter (quartz) was additionally applied to monitor the melt smoothly. As a result of applying the newly manufactured CCTV to the demonstration test, smooth monitoring ability was confirmed even at normal operating temperature (above 1,500°C). Through this facility improvement, the operation convenience of the plasma torch melting facility has been secured, and it is expected that it will be able to operate stably during long-term continuous operation in the future.

Currently, KHNP-CRI has developed 100 kW plasma torch melting facility to reduce the amount of radioactive waste in nuclear power plant. Plasma torch melting technology uses electric arc phenomena like lightning to melt the target material at a high temperature of about 1,600°C. The technology is applicable to treatment for various types of waste such as combustible, non-combustible and mixed wastes. The volume reduction ratio by the technology is respectively expected to be about 1/60 of combustible wastes and about 1/5 for non-combustible wastes. It is important to discharge the melt without problems in the melting technology. In general, molten slag has properties such as high viscosity and quick solidification. Because of the properties, when discharging into slag container, the final product is accumulated like a mountain. To improve this problem, there is three suggestions; 1) rotation of the slag container, 2) vibration of the slag container, and 3) heating of the slag container.

The homogeneity of radioactive spent ion exchange resins (IERs) distribution inside waste form is one of the important characteristics for acceptance of waste forms in long-term storage because heterogenous immobilization can lead to the poor structural stability of waste form. In this study, the homogeneity of metakaolin-based geopolymer waste form containing simulant IERs was evaluated using a laser-induced breakdown spectroscopy (LIBS) and statistical approach. The cation-anion mixed IERs (IRN150) were used to prepare the simulant spent IERs contaminated by non-radioactive Cs, Fe, Cr, Mn, Ni, Co, and Sr (0.44, 8.03, 6.22, 4.21, 4.66, 0.48, and 0.90 mg/g-dried IER, respectively). The K2SiO3 solution to metakaolin ratio was kept constant at 1.2 and spent IERs loading was 5wt%. For the synthesis of homogeneous geopolymer waste form, spent IERs were mixed with K2SiO3 solution and metakaolin first, and then the fresh mixture slurry was poured into plastic molds (diameter: 2.9 cm and height: 6.0 cm). The heterogeneous geopolymer waste form was also fabricated by stacking two kinds of mixtures (8wt% IERs loading in bottom and 2wt% in top) in one mold. Geopolymers were cured for 7d (1d at room temperature and 6d at 60°C). The hardened geopolymers were cut into top, middle, and bottom parts. The LIBS spectra and intensities for Cs were obtained from the top and bottom of each part. Cs was selected for target nuclide because of its good sensitivity for measurement. Shapiro-Wilk test was performed to determine the normality of LIBS data, and it revealed that data from the homogeneous sample is normal distribution (p-value = 0.9246, if p-value is higher than 0.05, it is considered as normal distribution). However, data from the heterogeneous sample showed abnormal distribution (p-value = 7.765×10-8). The coefficient of variation (CoV) was also calculated to examine the dispersion of data. It was 31.3% and 51.8% from homogeneous and heterogeneous samples, respectively. These results suggest that LIBS analysis and statistical approaches can be used to evaluate the homogeneity of waste forms for the acceptance criterion in repositories.

Korea currently has two permanent shutdown Nuclear Power Plants (NPPs), and the decommissioning project is expected to begin soon, starting with the first commercial NPP. The decommissioning project will eventually be the disposal of radioactive waste in the final stage of the work, and in that respect, proper tracking and history management should be well established in the management of waste. This is in line with the guidelines that regulatory agencies should also properly manage radioactive waste. Therefore, this study intends to examine the factors that should be considered in terms of tracking and management of radioactive waste in decommissioning nuclear facilities. The starting and final point of tracking radioactive waste generated during decommissioning is the physical inventory of the current as-is state and the final container. In this respect, the tracking of waste starts from the beginning of the dismantling operation. Thus, at the stage of approval of the decommissioning work, it may begin with an ID scheme, such as the functional location in operation for the target System, Structure, and Components (SSCs). As the dismantling work progresses, SSCs will be classified by nature and radiological level, which will be placed in containers in small packaging units. At this time, the small package should be given an ID. After that, the dismantling work leads to the treatment of waste, which involves a series of operations such as cutting, decomposition, melting, and decontamination. Each step in which these tasks are performed will be placed in a container, and ID assignment is also required. Until now, the small packaging container is for transfer after each treatment, and it is placed in the storage container in the final stage, at which time the storage container also gives a unique ID. Considerations for follow-up management were reviewed assuming solid waste, which is the majority of dismantled radioactive waste considered in this study. The ID system should be prepared from the start of the dismantling work, ID generation of the small transporting container and ID generation of the final disposal container during the intermediate waste treatment process, and each ID generation of the previous stage should be linked to each generation stage. In addition, each ID must be generated, and the definition of the grant scheme and attributes is required.

In nuclear decommissioning projects, past and present projects in the world, an important area to be managed is waste management. The management of waste should be done with various aspects of consideration in mind from the moment it occurs from the cutting and dismantling of Systems, Structures, and Components (SSCs). Therefore, this study aims to discuss the disposition considerations for the efficient management of low and very low level waste that is expected to be generated in large quantities and to examine its applicability to domestic nuclear facilities. As for waste management, radioactive wastes begin to be generated when SSCs are dismantled, so waste management should be carried out as a result of dismantling activities. In addition, the waste is stored in the final disposal container and transferred to the storage or disposable facilities. In order to store in the final disposal container or transfer container, it will have to be classified by radioactive level. From the perspective of waste classification, wastes below the low level can be divided into low levels, very low levels, and clearance in Korea. Therefore, as an important point of waste management, when SSCs are dismantled, the work process must be carried out until the final disposal in accordance with the disposition strategy based on the waste classification. As a disposition strategy, the process presented by the IAEA can be referred. The materials to be dismantled for the first time are largely divided into radioactive and suspected radioactive materials. After going through the dismantling process, three criteria are considered to satisfy the disposition option: unconditional release criterion, conditional criterion, and radioactive waste. The types of waste below the final low level are classified into two types as unconditional, two types as conditional, and low and very low levels. In this study, six disposition options are reviewed, including unconditioned reuse and disposal, conditional reuse and disposal, and disposal of VLLW and VLW. Options for radioactive waste may be subject to operational criteria and may need to be supplemented in terms of the acceptance criteria in the repository. In the case of the conditional option, the clearance criterion can be applied, but considering the decommissioning characteristics, it is an option that can be used for nuclear industry, and specific reuse scenarios should be supplemented through discussions with the regulatory agency. In addition, it seems that the unconditional option needs to establish a corresponding criterion.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

In the Kori-1 radioactive waste storage, the concentrated waste and spent resin drums generated in the past are repacked and stored in large concrete drums. In order to dispose of radioactive waste generated before the establishment of the waste acceptance criteria, it is necessary to develop a large concrete drum treatment and waste treatment process to evaluate disposal suitability and secure technology that meets the latest technical standards. In addition, for worker safety and waste reduction, it is important to develop secondary waste treatment technology generated during waste treatment. In this study, the types and characteristics of secondary wastes that can be generated when large concrete drums are decommissioned were investigated. In addition, considering the characteristics of possible secondary wastes, suitable treatment methods and characteristic evaluations were analyzed. We plan to develop an optimal process for secondary waste treatment in consideration of on-site work space, economic feasibility, and safety.

Thermal treatment, such as combustion, is the most effective way to solve the spatial problem of radioactive waste disposal. Existing incineration technology has the problem of discharging harmful pollutants (CO2 and dioxin, etc.) into the environment. Therefore, it was evaluate the validity of the thermal treatment process that can reduce the volume of dry active waste (DAW) in an eco-friendly. In addition, the stability of the alternative incineration process under development was evaluated by evaluating the emission of harmful pollutants to the environment during the thermal treatment process. We selected 14 samples identical to those discarded by each nuclear power plant (Kori, Saeul, Wolsong, Hanbit, Hanul). And EA (Elemental Analysis) analysis was performed on each sample. As a result, excluded samples containing wastes containing POPs (Persistent Organic Pollutants) such as PCBs (Polychlorinated Biphenyls), which could generate harmful pollutants during thermal treatment, and halogenated organic wastes such as PVC (Polyvinyl Chloride). In addition, the thermal treatment conditions for the four DAWs were derived by Thermogravimetric Analysis/Differential Thermal Analysis (TG/DTA) analysis. At this time, Py-GC/MS analysis was performed at the temperature at which each waste causes thermal decomposition (cotton is 437°C, paper is 562°C, latex glove is 430°C, plastic bag is 485°C). As a result of analyzing the exhaust gas produced during thermal decomposition, about 77.0% of the cotton was Benzoic acid series, the paper was 41.1% Glucopyranose series, and 15.8% hydroxy acetaldehyde. Latex glove was identified to be 45.9% and 19.2% for Limonene and 2-methyl-1, 3-Butadiene, and for plastic bags, Octacosanol and 2-octyl-1-Dodecanol were 38.8% and 15.2%. In addition, it was confirmed that dioxin and harmful heavy metals, which are discussed as environmental risks, were not detected in all samples.

A lot of solid wastes are generated when nuclear power plant is dismantled, and a lot of treatment costs and optimal waste treatment technologies are required to treat the generated solid wastes. Currently, there is no optimized reduction and solidification technology for each characteristics of radioactive dismantling waste, so the customized treatment technology for each waste is required to respond actively to this issue. This paper shows the evaluation results of molding and sintering characteristics using preliminary sample to derive operational characteristics and improvements for powder mixing device, molding device, and sintering device manufactured for solidification of dispersible radioactive waste. Zeolite was used as a preliminary sample for performing basic operation characteristics evaluation of each unit device. First of all, the basic operation characteristics of the powder mixing device was evaluated by analyzing the sample distribution, mixing degree, and tap density. It was confirmed that the preliminary sample was well mixed in all areas of the cylinder where the mixing was performed. In the tap density analysis, the increase effect of the volume reduction of the sample was confirmed according to the increase of the RPM speed (up to 2000 RPM). Since the particle size of zeolite sample is very small (nanometer size), the particular consistency of the change of average particle size with RPM speed couldn’t be confirmed, but the uniform of particle size distribution was confirmed with RPM speed size. The basic operation characteristics of the molding device was evaluated for each mold size (ID30, ID50, ID100) according to the moisture content (0-20%) and the molding pressure condition (25-200 MPa) for the preliminary sample. In the characteristics evaluation of the sintered body, the strength of the sintered body was much higher than that of the molded body. However, it was confirmed that as moisture evaporated during the sintering process according to the moisture content contained in the molded body, the swelling occurred in the sintered body due to vapor pressure, and this caused cracks in the longitudinal or transverse direction inside and outside the sintered body. Therefore, optimal moisture content conditions for sintering should be derived. In conclusion, if the operation characteristics and improvements of powder mixing, molding and sintering devices derived from this study are reflected and improved, it is judged that it is possible to derive the optimal process for solidification of dispersive radioactive wastes.

In underground repository environments, various types of engineered barriers are installed to hinder the mobility of radionuclides. Cement admixtures, especially used to improve workability for concrete, are composed of fairly high organic molecules and have a dispersing effect through bonding with the C-S-H of the concrete. Previous studies have shown that complex-forming organics like EDTA, NTA, and ISA have a significant effect on the mobility of radionuclides, but the studies on the behavior and stability of combined complexes in hydrated cement are lacking. So, we selected a commonly used polycarboxylic-ester (PCE) type cement admixture and stable Co as a surrogate of Co-60 to perform desorption experiments from hydrated cement containing the admixture. Radioactive Co is known to be a common contaminant in nuclear fission and medical facilities and considered to exist as a relatively stable phase in repositories. In addition, the evaluation of cobalt can be a standard of safety issue for other radionuclides with the presence of cement admixture in repository. In this study, cement samples were prepared at water/cement ratio of 0.55 and cured for 28 days at 23-25°C and at least 80% of humidity with varying cement admixtures of 0.0, 0.1, and 2.0wt%. To evaluate the stability of cobalt in the weathered cement, a 0.001 M HCl solution was used to simulate cement weathering conditions on a hot plate at 60°C for 1 day using a solid/liquid ratio of 1:100. Degree of weathering was confirmed using XRD analysis. The adsorption experiments were performed by adding 0.0042 mmol of cobalt (CoCl2, Sigma-Aldrich, anhydrous ≥ 98.0%) to the weathered cement for 3 days using a platform shaker at 200 rpm, and the supernatant was separated using a syringe filter (<0.20 um) before ICP-MS analysis to determine the amount of Co adsorption. Cobalt desorption was tested for the Co-adsorbed cement using 0.019 mmol of calcium (Ca(NO3)2·4H2O, Sigma-Aldrich, 99%) for 3 hours to 14 days. The results showed that adsorbed cobalt with and without cement admixture was stably bound to cement, and did not increase any noticeable Co release by 2.0wt% PCE admixture. However, additional experiments using varying contents of PCE and other admixtures should be conducted to provide a standard for assessing the safety of cement admixtures in repositories.

In Korea, many characteristic component facilities and technologies in general experimental areas for non-radiative materials are owned by industry-academia research. Still, no characteristic analysis test technology has been developed for large, intermediate-level decommissioning waste emitted by neutron irradiation. Since Korea plans to decommission nuclear power plants in 2027, securing analysis technology for intermediate-level decommissioning waste is essential. Accordingly, the Korea Research Institute of Decommissioning (KRID) plans to secure an infrastructure (hot cell) to analyze the characteristics of intermediate-level dismantled waste. Afterward, we intend to stably dispose of the waste generated while decommissioning the current Gori Unit 1/Wolseong Unit 1 using the intermediatelevel dedicated hot cell. It aims to secure high-dose/high-radiation decommissioning waste handling technology through intermediate-level hot cells for the first time in Korea, supports domestic nucleardecommissioning projects, and secure and validate procedures related to material characteristics and nuclide analysis of intermediate-level waste. Furthermore, research on intermediate-level radioactive materials is expected to be carried out in cooperation with schools and research institutes.

With the recent concern regarding cellulose enhancing radionuclide mobility upon its degradation to ISA, disposal of cellulosic wastes is being held off until the disposal safety is vindicated. Thus, a rational assessment should be conducted, applying an appropriate cellulose degradation model considering the disposal environment and cellulose degradation mechanisms. In this paper cellulose degradation mechanisms and the disposal environment are studied to propose the best-suitable cellulose degradation model for the domestic 1st phase repository. For the cellulose to readily degrade, the pH should be greater than 12.5. As in the case of SKB, 1BLA is excluded from the safety assessment because the pH of 1BLA remains below 12.5. Furthermore, despite cellulose degradation occurring, it does not always produce ISA. At low Ca2+ concentration, the ISA yield rate is around 25%, but at high Ca2+ concentration, the ISA yield rate increases up to 90%. Thus, for the cellulose to be a major concern, both pH and Ca2+ concentration conditions must be satisfied. To satisfy both conditions, the cement hydration must be in 2nd phase, when the porewater pH remains around 12.5 and a significant amount of Ca2+ ion is leaching out from the cement. However, according to the safety evaluation and domestic research, 2nd phase of cement hydration for silo concrete would achieve a pH of around 12.4, dissatisfying cellulose degradation condition like in 1BLA. Thus, cellulose degradation would be unlikely to occur in the domestic 1st phase repository. To derive waste acceptance criteria, a quantitative evaluation should be conducted, conservatively assuming cellulose is degraded. To conduct a safety evaluation, an appropriate degradation model should be applied to determine the degradation rate of cellulose. According to overseas research, despite the mid-chain scission being yet to be seen in the experiments, the degradation model considering mid-chain scission is applied, resulting in an almost 100% degradation rate. The model is selected because the repositories are backfilled with cement, achieving a pH greater than 13, so extensive degradation is reasonably conservative. However, under the domestic disposal condition, where cellulose degradation is unlikely to occur, applying such model would be excessively conservative. Thus, the peeling and stopping model derived by Van Loon and Haas, which suggests 10~25% degradation rate, is reasonably conservative. Based on this model, cellulose would not be a major concern in the domestic 1st phase repository. In the future, this study could be used as fundamental data for planning waste acceptance criteria.

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.

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

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.

Bentonite is a widely used buffer material in high-level radioactive waste repositories due to its favorable properties, including its ability to swell and low permeability. Bentonite buffers play an important role in safe disposal by providing a low permeability barrier and preventing radionuclides migration into the surrounding rock. However, the long-term performance of the bentonite buffer is still an area of research, and one of the main concerns is the erosion of the buffer due to swelling and groundwater flow. Erosion of the bentonite buffer can have a significant impact on repository safety by reducing the integrity of the buffer and forming colloids that can transport radionuclides through groundwater, potentially increasing the risk of radionuclide migration. Therefore, understanding the mechanisms and factors that influence the erosion of the bentonite buffer is critical to the safety assessment of high-level radioactive waste repositories. In this study, we attempted to develop the bentonite buffer erosion model using Adaptive Processbased total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). First, the erosion phenomenon was divided into two stages: bentonite buffer penetration into rock fractures and colloid formation. As an initial step in the development of the buffer erosion model, a bentonite buffer intrusion into the fracture and consequent degradation of buffer property were considered. For this purpose, a tworegion model based on the dynamic bentonite diffusion model was adopted which is one of the methods for simulating bentonite buffer intrusion. And, it was assumed that the buffer properties, such as density, porosity and permeability, thermal conductivity, modulus of elasticity, and mechanical strength, are degraded as the buffer erodes. The bentonite buffer degradation model developed in this study will serve as a foundation for the comprehensive buffer erosion model, in conjunction with the colloidal formation model in the future.

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.

DC plasma torch is reported as a technology that can be treated regardless of waste types because it can select transfer or non-transfer operation modes depending on the electrical conductivity of waste. Thanks to this characteristic, countries that operate nuclear power plants such as Switzerland, Japan, and Taiwan have developed high-power DC plasma torch to dispose of radioactive waste. And also in korea, a plasma torch is being developed to dispose of radioactive waste. This study was conducted to investigate the characteristics of the reverse polarity plasma torch according to the conditions of interelectrodes. The inter-electrodes of plasma torch used in the study was designed to be 25, 37 mm in diameter and 180 to 400 mm in length. As a result of the test, it was confirmed that the smaller the diameter and the longer the length of the inter-electrode, the more advantageous it was to achieve a high output power. And it was confirmed that the power torch would be 500 kW when the diameter of the inter-electrode was 25 mm, the length was 400 mm, and the current was 500 A.

It is important that the plasma torch used in the waste treatment field has a high output to increase throughput. In order to increase the output of the plasma torch, there is a method of increasing the current or extending the length of the plasma arc. Among these methods, high power can be easily achieved simply by increasing current, but it is difficult to ensure electrode life. Therefore, it is necessary to check the appropriate current and arc length conditions to achieve high power and stable operation. In this paper, the power performance according to the arc length, current, and operation mode was confirmed in the transfer mode plasma torch. The test conditions are the distance (arc length) between the plasma torch and the external electrode was set to 5-180 mm, and the current was set to be in the range of 90-460 A. As a result of the test, it was confirmed that the reverse polarity operation had a maximum output of 159 kW depending on the arc length and current, and the positive polarity operation had a maximum output of 138 kW. Through this result, it was confirmed that the arc length had an effect on increasing the output, and that the reverse polarity operation had a longer arc than the positive polarity operation.

Depending on the type of waste, DC plasma torch uses a transfer type operation for conductive waste and a non-transfer type operation for non-conductive waste. The transfer mode plasma torch can secure high throughput because the arc directly contacts the object and has high thermal efficiency. However, since the non-transfer mode does not have a higher thermal efficiency than the transfer mode, higher output is required to secure high throughput. A method of increasing the output of the plasma torch is increasing the current or extending the length of the plasma arc. However, the method of increasing the current affects the life of the electrode, and there is a limit to extending the arc length in the positive polarity plasma torch. Therefore, it is effective to design the plasma torch with reverse polarity to secure life and extend the arc length. In the reverse polarity plasma torch, the front electrode serves as the cathode, and the cathode point is not easy to control compared to the anode point, which may cause abnormal arcing and damage the plasma torch. This paper was conducted to investigate the conditions for securing the safety of these non-transferable reverse polarity plasma torch. The plasma torch is designed to have an output of 100 kW or less and to use the detachable nozzle to control the cathode point. The test showed that the shape of the nozzle prevented the cathode point moving outside of plasma torch and the excessive extension of the arc. Thanks to this, it was confirmed that plasma could be stably formed and abnormal arcing could also be prevented.