Dry active waste (DAW) contains substantial amount of cellulose related materials. The DAW are usually classified as low and/or very low-level waste. In Korea, three types of disposal facilities have been considered: silo, engineering barrier, and land-fill. Currently, only the silo type disposal facility is in operation. Around 27 thousand drums were disposed in silo. Massive amount of cement concrete is used in construction of silo. The ground waste, which flow through the concrete structure, shows higher pH than as it is. It is generally known that the pH of silo is ~12.47 in Korea, when considering construction material, filling material, and property of ground water. It is expected that the cellulose in DAW will be partially transformed to isosaccharinic acid (ISA). It is generally accepted that the ISA plays a negative role in safety analysis of disposal facility by stimulation of specific nuclides. Various factors affect the degradation of cellulose containing radioactive waste, such as degree of polymerization, pH of disposal condition, interaction between concrete structure and ground water, etc. In this paper, the disposal safety analysis of cellulose containing radioactive, usually paper, cotton, wood, etc., are studied. The degradation of cellulose with respect to degree of polymerization, pH of neighboring water, filling material of silo, etc. are reviewed. Based on the review results, it is reasonable to conclude that the substantial amount of DAW could be disposed in silo.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. In addition, according to the level of radioactivity, it can be divided into high level, intermediate level, low level, and clearance level waste. In the case of solid radioactive waste, it is necessary to secure disposal suitability in order to deliver it to a disposal facility, so safe and efficient treatment of a large amount of radioactive waste generated during decommissioning is one of the most important issues. For the treatment of radioactive waste generated during decommissioning, technologies in various fields such as cutting, decontamination, melting, measurement, and packaging are required. Therefore, this study intends to present and application plan for decommissioning domestic nuclear power plants through overseas case studies for the treatment of radioactive waste expected to occur during nuclear power plant decommissioning.

As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

After the Fukushima accident, significant amount of radioactively contaminated waste has been generated with 50~250 m3/day and stored in tanks of the Fukushima Daiichi nuclear power plant site. The contaminated water is treated by various treatment facility such as KURION, SARRY, Reverse Osmosis, and ALPS to remove 62 radioactive nuclides except H-3. For the contaminated water treatment process, massive secondary wastes such as sludge, spent adsorbent, and so on as by-product are being generated by the facilities. In Japan, to treat the secondary wastes, melting technologies such as GeoMelt, In-can vitrification and Cold Crucible Induction Melting vitrification are considered as a candidate technologies. In this study, the technologies were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for treatment of the secondary wastes.

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.

Domestic nuclear power plants developed the scaling factors for the radioactive waste generated from 2004 to 2008 for the first time. Afterwards, the effectiveness of continuous application of the scaling factors have been evaluated for newly generated radioactive waste over the past two years. It was confirmed that most of the initially developed scale factors were effective within a factor of 10 principle, which is an acceptable range. The scaling factors were updated using the analysis data base from 2004 to 2016. A scaling factor refers to the calculated abundance ratio between Key (Easy-to- Measure) and DTM (Difficult-to-Measure) nuclide at the time of generation of radioactive waste based on the source term in the reactor of an operating nuclear power plant. The effectiveness of continuous application of scaling factors can be evaluated at regular intervals regardless of operation status or when there are events that change scaling factors during nuclear power plant operation, such as zinc injection, large-scale facility replacement, and long-term shutdown etc. Even in the case of a permanently shut down nuclear power plant in which fuel is withdrawn from the reactor and generation of new nuclides by nuclear fission and radiation has stopped, periodic verification is conducted to confirm whether the scaling factor developed before permanent shutdown can be effectively applied to the radioactive waste generated after permanent shutdown. However, depending on the nuclear power plant decommissioning strategy or conditions, the period of permanent shutdown prior to decommissioning can be very long, so preparations are needed to ensure the appropriateness of scaling factor operation. In the case of domestic nuclear power plants, Kori Unit 1, a light water reactor, was permanently shut down in June 2017, and as a heavy water reactor nuclear power plant, the permanent shutdown of Wolseong Unit 1 was finally decided in December 2019 after twists and turns including large-scale facility replacement and long-term shutdown. In this paper, we have predicted when the scaling factors will change significantly due to radioactive decay and the difference in halflife between the Key and DTM nuclides over time after permanent shutdown. We also have tried to find appropriate countermeasures for the operation of scaling factors during permanent shutdown period, such as updating scaling factors or applying correction factors.

A disposal of radioactive wastes is one of the critical issues in our society. Considering upcoming plans for dismantling of nuclear power plants, this problem is inevitable and should be discussed very carefully. There are variety of methods to handle with radioactive wastes, including Incineration, conventional gasification and plasma gasification. Among them, plasma gasification process is in the limelight due to its eco-friendly & stable operation, and large volume reduction effects. However, a fatal disadvantage is that it consumes more electric power than other methods, this leaves us a question of whether this process is indeed economical. Within the scope of this paper, I would like to introduce 4 cases which plasma facilities were evaluated economically in worldwide, and reach the conclusion on the economic feasibility of plasma process.

Decommissioning of Nuclear Power Plant (NPP) projects in South Korea starts with permanent shutdown of Kori unit 1 and Wolsung unit 1. It is important to establish a treatment and disposal method for radioactive waste generated during the decommissioning of the nuclear power plants. Large quantities of the wastes during decommissioning of NPP are generated in a short period of time and the wastes have various types and characteristics. For efficient decommissioning of NPP process, the radioactive waste is classified by types and each treatment method and packaging concept is presented respectively in this paper. Radioactive waste generated during decommissioning of NPP is classified into reactor vessel, reactor internals, metals, Dry Active Waste (DAW), concreate, spent fuel storage rack, spent resin and spent filter, etc., and the packaging concept for each type should be established to meet the waste acceptance criteria. Major waste acceptance criteria requirements include nuclides concentration, filling rate, free water, surface radiation does rate and weight. Radioactive waste containers can be classified into packaging containers, transport containers, and disposal containers. The packaging container is used to contain, transport, and store radioactive waste within the radiation control area, and a control number has been assigned as a radioactive waste drum after the final treatment has been completed. The transport container is used for transporting radioactive waste filled-containers from a radiation control area through an uncontrolled area. In this paper, the concept of disposal of dismantled radioactive waste and packaging methods were reviewed in comprehensive consideration of domestic radioactive waste transport and storage regulations, permanent disposal environment, and development status of large containers.

Spent filters contained in drums of radioactive waste generated from nuclear power plants are contaminated with various radioactive isotopes due to their use in various water purification processes in the system. Radiation doses from the spent filters can vary from low to high levels. To dispose of drums containing spent filters as radioactive waste, the inventory of radioactive isotopes in the filters must be determined. Two methods for determining the inventory are indirect measurement using scaling factors and direct analysis of filter samples. This study suggests a method to determine the appropriate sample size for each drum based on the number of filters stored in the drum, when direct analysis is used to determine the inventory of radioactive isotopes. In particular, Visual Sample Plan (PNNL) software’s Item Sampling function was used to calculate the sample size, considering the confidence level and minimum acceptable coverage rate. As a result, assuming that the number of filters packed per drum ranges from a minimum of 1 to a maximum of 30, the study suggests that a full inspection is required for drums containing 9 or fewer filters, while drums containing 10 filters should be sampled with 9 samples, 11 filters with 10 samples, 12-13 filters with 11 samples, 14-16 filters with 12 samples, 19-22 filters with 14 samples, 23-26 filters with 15 samples, and 27-30 filters with 16 samples.

KHNP-CRI has developed small-capacity and Mega-Watt Class PTM (Plasma Torch Melter) for the purpose of reducing the volume of radioactive waste and immobilizing or solidifying radioactive materials. About 1 MW PTM is a treatment technology that operates a plasma torch and puts drumshaped waste into a melter and radioactive waste in the form of slag is discharged into a waste container. The small-capacity PTM is a treatment technology that operates a plasma torch and puts small amounts of radioactive waste by directly putting it into the melter through a waste input machine. Mega-Watt Class PTM was able to inject radioactive waste in drums, so it was disposed of without backloging. On the other hand, The small-capacity PTM put radioactive waste without a package, and the waste input was blocked. If even small-capacity PTM put radioactive waste in the form of small packages such as drums, it is expected that various types of radioactive waste can be processed for a long time. Packaging also reduces the risk of radioactive contamination.

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.

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.

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.

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.

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.

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.

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.

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.

Plasma torch melting has been considered as a promising treatment technology for radioactive waste generated by nuclear power plants. The IAEA reported in 2006, the plasma melting technology could be treated regardless of the type of radioactive wastes such as combustible, non-combustible and liquid. Also, the technology has the advantage of being an eco-friendly technology. It emits less harmful gases such as NOx, SOx, HCl and CO because it does not use fossil fuels. In KHNP CRI, the plasma torch melting system was developed as the new radioactive waste treatment technology. In this study, to evaluate the long-term integrity of the new facility, a demonstration test with concrete as a simulant was carried out for about 3 days. For the 3 days, the evaluation was conducted in view of abnormal shutdown, soundness of waste feeding device, electrode consumption, and so on.

The treatment of waste generated during operation as a part of preparation for decommissioning is coming to the fore as a pending issue. Non-fuel waste stored in the spent fuel pool (SFP) of PWRs in Korea includes Dummy fuel, damaged fuel rod storage container, reactor vessel specimen cask, spent in-core instrumentation, spent control element assemblies, spent neutron source assemblies, burnable poison rods, etc. In order to treat such waste, it is necessary to classify radioactive waste level and analyze kinds of nuclide in accordance with legal requirements. In order to solve the problem, the items that KHNP-CRI is trying to conduct like followings. First, KHNP-CRI will identify the current status of non-fuel waste stored in the SFP of all domestic nuclear power plants. In order to consider the treatment of non-fuel waste, it is essential to know what kind of items and how many items are stored in the SFP. Second, to identify the dimension and characteristics of non-fuel waste stored in the SFP would be conducted. The configuration of non-fuel waste is important information to handle them. Third, the way to handle non-fuel waste would be deduced including analysis of their dimension, whether the equipment should be developed to handle each kind of non-fuel waste or not, how to transport them. In order to classify radioactive waste level and analyze the nuclide for the non-fuel waste, handling tools and the cask to transport them into the facility which nuclide analysis is able to be performed would be required. Fourth, the nuclide analysis technology would be identified. Also, domestic holding technology would be identified and which technology should be developed to classify the radioactive waste level for the non-fuel waste would be deduced. This preliminary study will provide KHNP-CRI with the insight for the nuclide analysis technology and future work which is following action for the non-fuel waste. Based on the result of above preliminary study, the feasibility of the research for the treatment of non-fuel waste would be evaluated and research plan would be established. In conclusion, the treatment of non-fuel waste stored in the spent fuel pool of domestic PWR should be considered to prepare the decommissioning. KHNP-CRI will identify the quantity, the dimension and kinds of non-fuel waste in the SFP of domestic PWR. Also, the various nuclide analysis technology would be identified and the technology which should be developed would be defined through this preliminary study.